WorldWideScience

Sample records for experimental gas cooled reactor

  1. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  2. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    coolants. The purpose of the high temperature helium loop (HTHL) is to simulate technical and chemical conditions of VHTR's coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes also the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.

  3. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  4. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  5. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  6. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

  7. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  8. Cooling performance of a water-cooling panel system for modular high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Shoji; Suzuki, Kunihiko; Inagaki, Yoshiyuki; Sudo, Yukio [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1995-12-31

    Experiments on a water cooling panel system were performed to investigate its heat removal performance and the temperature distribution of components for a modular high-temperature gas-cooled reactor (MHTGR). The analytical code THANPACST2 was applied to analyze the experimental results to verify the validity of the analytical method and the model.

  9. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  10. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...... extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between...

  11. Gas cool reactor operation in the UK. The present position

    Energy Technology Data Exchange (ETDEWEB)

    Marsden, B.J. [AEA Technology, Risley, Warrington, Cheshire (United Kingdom)

    1998-09-01

    During 1996 there was a major reorganisation of the UK Nuclear Industry. The Advanced Gas Cooled Reactors (AGRs) and Pressurised Water reactor (PWR) operated by Nuclear Electric and Scottish Nuclear was privatised under a new company called British Energy. The Magnox reactors are now operated by a public utility named Magnox Electric, which is currently being merged with British Nuclear Fuels (BNFL). The old UKAEA was split into two parts; a public company, which has maintained the name UKAEA that is responsible for managing all their nuclear liabilities and a technical/consultancy company which is privately owned under the name of AEA Technology. Most of the Magnox and AGRs have continued to operate well with high availability factors. Decommissioning programmes have continued to expand and decommissioning costs have reduced below predicted levels. The industry is maintaining its safety research programme on gas cooled reactors under the direction of the HSE

  12. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  13. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  14. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  15. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  16. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  17. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    Science.gov (United States)

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

  18. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  19. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  20. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  1. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  2. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  3. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  4. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  5. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  6. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  7. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  8. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  9. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  10. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  11. Gas Reactor International Cooperative Program. Interim report: assessment of gas-cooled reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    The merits of introducing Pebble Bed Gas Reactors (PBRs) into the existing US electric generating sector are discussed. Information is presented concerning an economic model; nuclear fuel costs; capital cost targets; time comparison of nuclear power costs; introduction scenarios; domestic economic incentives; the selection of a discount rate for national energy supply studies; nuclear fuel cycle cost calculation code RAMMER; and PBR and HTGR fabrication and reprocessing costs.

  12. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all

  13. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    Science.gov (United States)

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  14. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  15. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  16. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  17. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  18. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  19. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    OpenAIRE

    Shutang Zhu; Ying Tang; Kun Xiao; Zuoyi Zhang

    2008-01-01

    This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR) technology with the supercritical (SC) steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR) reveal that the development of SCWR power plants still needs further research and develop...

  20. A thermodynamic approach for advanced fuels of gas-cooled reactors

    Science.gov (United States)

    Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  1. A thermodynamic approach for advanced fuels of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gueneau, C. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France)]. E-mail: cgueneau@cea.fr; Chatain, S. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Gosse, S. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Rado, C. [DEN/DTEC/STCF - CEA Valrho, 26702 Pierrelatte cedex (France); Rapaud, O. [DEN/DTEC/STCF - CEA Valrho, 26702 Pierrelatte cedex (France); Lechelle, J. [DEN/DEC/SPUA - CEA Cadarache, 13108 Saint-Paul Lez Durance cedex (France); Dumas, J.C. [DEN/DEC/SESC - CEA Cadarache, 13108 Saint-Paul Lez Durance cedex (France); Chatillon, C. [LTPCM - UMR5614, ENSEEG BP75 Grenoble, 38402 Saint-Martin d' Heres cedex (France)

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO{sub 2} gas formation during the chemical interaction of [UO{sub 2{+-}}{sub x}/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  2. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  3. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  4. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  5. Power cycle assessment of nuclear high temperature gas-cooled reactors

    OpenAIRE

    Herranz, L.E.; Linares, J.I.; Moratilla, B.Y.

    2009-01-01

    Power cycle assessment of nuclear high temperature gas-cooled reactors correspondance: Corresponding author. Tel.: +34 91 346 62 36; fax: +34 91 346 62 33. (Herranz, L.E.) (Herranz, L.E.) Unit of Nuclear Safety Research (CIEMAT) Avda. Complutense--> , 22 - 28040 Madrid - Spain--> - (Herranz, L.E.) Unit of Nuclear Safety Research (CIEMAT) Avda. Complutense--> , 22 - 28040 Madrid - Spain--...

  6. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  7. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  8. Accident Analysis Simulation in Modular 300MWt Gas Cooled Fast Reactor

    Science.gov (United States)

    Zaki, Su'ud

    2017-01-01

    Safety analysis of 300MWt helium gas cooled long-life fast reactors has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 300 MWt He gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. GCFR relatively has hard spectrum so it has relatively small Doppler coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.002dk/k. In addition the steam generator design has also consider excess power during severe UTOP case..

  9. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  10. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  11. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  12. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  13. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  14. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Drajat, R. Z.; Su' ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

  15. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  16. Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

    Science.gov (United States)

    Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

    2003-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

  17. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    Science.gov (United States)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

    2002-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  18. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  19. Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wu Zongxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Sun Yuliang [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Li Fu [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)]. E-mail: lifu@tsinghua.edu.cn

    2006-03-15

    The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.

  20. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  1. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  2. The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

    Energy Technology Data Exchange (ETDEWEB)

    Gotschall, H.L. (Gas-Cooled Reactor Associates, San Diego, CA (United States))

    1994-01-01

    The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership.

  3. A modular gas-cooled cermet reactor system for planetary base power

    Science.gov (United States)

    Jahshan, Salim N.; Borkowski, Jeffrey A.

    1993-01-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  4. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  5. Commercialization of modular high temperature gas-cooled reactors in the world

    Energy Technology Data Exchange (ETDEWEB)

    Hayakawa, Hitoshi; Okamoto, Futoshi; Ohhashi, Kazutaka [Fuji Electric Co. Ltd., Tokyo (Japan)

    2001-07-01

    The construction programs of the commercial high temperature gas-cooled reactors have been activated extraordinarily all over the world. This paper gives an overview of the three major programs, the South African PBMR project (US utility Exelon announced recently their plan to import PBMRs), the international GT-MHR project (by US DOE, General Atomics, MINATOM of Russian Federation, FRAMATOME ANP, Fuji Electric) and Chinese HTR-PM project. And the reasons why the utilities selected small modular HTGRs as next generation reactors, the superior characteristics of the small modular HTGRs for power generation plant and prospects of them are summarized and discussed. (author)

  6. Gas-cooled thorium reactor with fuel block of the unified design

    Directory of Open Access Journals (Sweden)

    I.V. Shamanin

    2015-11-01

    Analysis of information materials pertaining to the use of thorium as fuel element in rector facilities of the new generation and of its future potential was performed in the present study. Results of the first phase of neutronics studies of 3D model of high-temperatures gas-cooled reactor facility on the basis of unified design of the fuel block are presented. Calculation 3D model was developed using the software code of the MCU-5 series. Several optimal configurations of the reactor core were selected according to the results of comparison of neutronics characteristics of the examined options for the purpose of development of small-size modular nuclear power installations with power up to 60MW. Results of calculations of reactivity margin of the reactor, neutron flux distribution and power density profiles are presented for the selected options of reactor core configuration.

  7. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  8. VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

    Directory of Open Access Journals (Sweden)

    NAM-IL TAK

    2013-11-01

    Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

  9. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  10. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  11. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  12. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  13. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    OpenAIRE

    S. Yamoah; E.H.K. Akaho; Nana G.A. Ayensu; M. Asamoah

    2012-01-01

    The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as ther...

  14. Heat transfer characteristics in depressurized LOFC accidents with a failure of the RCCS in a modular gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seyun; Ha, Sangjun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Tak, Namil; Lim, Hongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    A modular gas-cooled reactor has inherent safety characteristics with its large heat capacity and low power density of the core when compared with conventional light water reactors. The reactor cavity cooling system (RCCS) serves as an ultimate heat sink in a high temperature gas-cooled reactor and is a system for the removal of the decay and residual heat from the uninsulated reactor vessel to ensure a plant safety. To understand the inherent safety features of the designed reactor, analyses for the RCCS performance in various severe accident conditions are required. A depressurized loss of forced circulation (LOFC) accident was considered as an initiating condition. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS is considered in this study.

  15. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  16. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  17. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  18. Simulation and control of water-gas shift packed bed reactor with inter-stage cooling

    Science.gov (United States)

    Saw, S. Z.; Nandong, J.

    2016-03-01

    Water-Gas Shift Reaction (WGSR) has become one of the well-known pathways for H2 production in industries. The issue with WGSR is that it is kinetically favored at high temperatures but thermodynamically favored at low temperatures, thus requiring careful consideration in the control design in order to ensure that the temperature used does not deactivate the catalyst. This paper studies the effect of a reactor arrangement with an inter-stage cooling implemented in the packed bed reactor to look at its effect on outlet temperature. A mathematical model is developed based on one-dimensional heat and mass transfers which incorporate the intra-particle effects. It is shown that the placement of the inter-stage cooling and the outlet temperature exiting the inter-stage cooling have strong influence on the reaction conversion. Several control strategies are explored for the process. It is shown that a feedback- feedforward control strategy using Multi-scale Control (MSC) is effective to regulate the reactor temperature profile which is critical to maintaining the catalysts activity.

  19. Analysis of the gas diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor

    OpenAIRE

    Zhang, Z.; Gerwin, Helmut; Scherer, Winfried

    1993-01-01

    In order to simulate the diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor, a one-dimensional coupled diffusion-convection model has been established. In this analysis it is shown first, that experiments performed at the Japan Atomic Energy Research Institute (JAERI) have been recalculated successfully, thus validating the new model. Applying this model to the NACOK facility, now under construction at the Institute for Safety Researc...

  20. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  1. Design of Helium Brayton Cycle for Small Modular High Temperature Gas cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoon Han; Lee, Je Kyoung; Lee, Jeong Ik [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2012-05-15

    The small modular reactor (SMR) is gaining a lot of interest recently. Not only it can achieve better passive safety, but also it can be potentially utilized for the diverse applications to respond to the increasing global energy demands. As a part of the SMR development effort, SM-HTGR (Small Modular-High Temperature Gas-cooled Reactor), a 20MWth reactor is under development by the Korean Atomic Energy Research Institute (KAERI) for the complete passive safety, desalination and industrial process heat application. The Helium Brayton cycle is considered as a promising candidate for the SM-HTGR power conversion. The advantages of Helium Brayton cycles are: 1) helium is an inert gas that does not interact with structure material. 2) helium is chemically stable that helium Brayton cycle can be utilized under the high temperature circumstance. 3) higher thermal efficiency is achievable under higher outlet temperature range. Moreover, high temperature advantage can be utilized (reinforced) by diverting part of the heat for industrial process heat. This paper will discuss the progress on the helium power conversion cycle operating condition optimization by studying the sensitivity of the maximum pressure, pressure ratio and the component cooling on the total cycle efficiency

  2. Assessment of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    OpenAIRE

    Zhang, Z.; Dong, Y; Scherer, W

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemente...

  3. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  4. Benchmark problem for International Atomic Energy Agency (IAEA) coordinated research program (CRP) on gas-cooled reactor (GCR) afterheat Removal

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Shoji; Shiina, Yasuaki; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hishida, Makoto; Sudo, Yukio

    1997-12-31

    In IAEA CRP on `Heat Transport and Afterheat Removal for GCRs under Accident Conditions`, experimental data of the JAERI`s cooling panel test apparatus were selected as benchmark problems to verify the validity of computational codes for design and evaluation of the performance of heat transfer and temperature distribution of components in the cooling panel system of the HTGR. The test apparatus was composed of a pressure vessel (P.V) with 1m in diameter and 3m in height, containing heaters with the maximum heating rate of 100kW simulating decay heat, cooling panels surrounding the P.V and the reactor cavity occupied by air at the atmospheric pressure. Seven experimental data were established as benchmark problems to evaluate the effect of natural convection of superheated gas on temperature distribution of the P.V and the performance of heat transfer of both the water and the air cooling panel systems. The analytical code THANPACST2 was applied to analyze two benchmark problems to verify the validity of the analytical methods and models proposed. Under the conditions at helium gas pressure of 0.73MPa and temperature of 210degC in the P.V of the water cooling panel system, temperatures of the P.V were well estimated within the errors of -14% to +27% compared with the experimental data. The analyses indicated that the heat transferred to the cooling panel was 11.4% less than the experimental value and the heat transferred by thermal radiation was 74.4% of the total heat input. (author)

  5. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  6. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  7. Transient analysis of nuclear graphite oxidation for high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Wei, E-mail: wxu12@mails.tsinghua.edu.cn; Shi, Lei; Zheng, Yanhua

    2016-09-15

    Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.

  8. Critical evaluation of high-temperature gas-cooled reactors applicable to coal conversion

    Energy Technology Data Exchange (ETDEWEB)

    Spiewak, I.; Jones, J.E. Jr.; Rittenhouse, P.L.; DeStefano, J.R.; Delene, J.G.

    1975-12-01

    A critical review is presented of the technology and costs of very high-temperature gas-cooled reactors (VHTRs) applicable to nuclear coal conversion. Coal conversion processes suitable for coupling to reactors are described. Vendor concepts of the VHTR are summarized. The materials requirements as a function of process temperature in the range 1400 to 2000/sup 0/F are analyzed. Components, environmental and safety factors, economics and nuclear fuel cycles are reviewed. It is concluded that process heat supply in the range 1400 to 1500/sup 0/F could be developed with a high degree of assurance. Process heat at 1600/sup 0/F would require considerably more materials development. While temperatures up to 2000/sup 0/F appear to be attainable, considerably more research and risk were involved. A demonstration plant would be required as a step in the commercialization of the VHTR.

  9. Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

    2016-09-15

    The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

  10. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

  11. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  12. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi [Inst. of Nuclear Energy Technology Tsinghua Univ., Beijing, BJ (China); Scherer, Winfried

    1997-12-31

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H{sub 2}O density increase rate in the reactor core is smaller than 0.3 kg/(m{sup 3}s). The liquid water vaporization in the steam generator and H{sub 2}O transport from the steam generator to the reactor core reduces the impulse of the H{sub 2}O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  13. Sensitivity studies of modular high-temperature gas-cooled reactor postulated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Syd [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6010 (United States)]. E-mail: sjb@ornl.gov

    2006-03-15

    The results of various accident scenario simulations for the two major modular high temperature gas-cooled reactor (HTGR) variants (prismatic and pebble bed cores) are presented. Sensitivity studies can help to quantify the uncertainty ranges of the predicted outcomes for variations in some of the more crucial system parameters, as well as for occurrences of equipment and/or operator failures or errors. In addition, sensitivity studies can guide further efforts in improving the design and determining where more (or less) R and D is appropriate. Both of the modular HTGR designs studied - the 400-MW(t) pebble bed modular reactor (PBMR, pebble) and the 600-MW(t) gas-turbine modular helium reactor (GT-MHR, prismatic) - show excellent accident prevention and mitigation capabilities because of their inherent passive safety features. The large thermal margins between operating and 'potential damage' temperatures, along with the typically very slow accident response times (approximate days to reach peak temperatures), tend to reduce concerns about uncertainties in the simulation models, the initiating events, and the equipment and operator responses.

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  16. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  17. Neutronic analysis stochastic distribution of fuel particles in Very High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Ji, Wei

    The Very High Temperature Gas-Cooled Reactor (VHTR) is a promising candidate for Generation IV designs due to its inherent safety, efficiency, and its proliferation-resistant and waste minimizing fuel cycle. A number of these advantages stem from its unique fuel design, consisting of a stochastic mixture of tiny (0.78mm diameter) microspheres with multiple coatings. However, the microsphere fuel regions represent point absorbers for resonance energy neutrons, resulting in the "double heterogeneity" for particle fuel. Special care must be taken to analyze this fuel in order to predict the spatial and spectral dependence of the neutron population in a steady-state reactor configuration. The challenges are considerable and resist brute force computation: there are over 1010 microspheres in a typical reactor configuration, with no hope of identifying individual microspheres in this stochastic mixture. Moreover, when individual microspheres "deplete" (e.g., burn the fissile isotope U-235 or transmute the fertile isotope U-238 (eventually) to Pu-239), the stochastic time-dependent nature of the depletion compounds the difficulty posed by the stochastic spatial mixture of the fuel, resulting in a prohibitive computational challenge. The goal of this research is to develop a methodology to analyze particle fuel randomly distributed in the reactor, accounting for the kernel absorptions as well as the stochastic depletion of the fuel mixture. This Ph.D. dissertation will address these challenges by developing a methodology for analyzing particle fuel that will be accurate enough to properly model stochastic particle fuel in both static and time-dependent configurations and yet be efficient enough to be used for routine analyses. This effort includes creation of a new physical model, development of a simulation algorithm, and application to real reactor configurations.

  18. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  19. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  20. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  1. Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Jianghai Li

    2017-01-01

    Full Text Available High-temperature gas-cooled reactors (HTGR can incorporate wireless sensor network (WSN technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs. This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA and support vector machine (SVM are developed.

  2. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  3. Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model

    Energy Technology Data Exchange (ETDEWEB)

    Scott, M.; Green, P.L. [Dynamics Research Group, Department of Mechanical Engineering, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); O’Driscoll, D. [EDF Energy, Barnett Way, Barnwood, Gloucester GL4 3RS (United Kingdom); Worden, K.; Sims, N.D. [Dynamics Research Group, Department of Mechanical Engineering, University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2016-08-15

    Highlights: • A model was made of the AGR control rod mechanism. • The aim was to better understand the performance when shutting down the reactor. • The model showed good agreement with test data. • Sensitivity analysis was carried out. • The results demonstrated the robustness of the system. - Abstract: A model has been made of the primary shutdown system of an Advanced Gas-cooled Reactor nuclear power station. The aim of this paper is to explore the use of sensitivity analysis techniques on this model. The two motivations for performing sensitivity analysis are to quantify how much individual uncertain parameters are responsible for the model output uncertainty, and to make predictions about what could happen if one or several parameters were to change. Global sensitivity analysis techniques were used based on Gaussian process emulation; the software package GEM-SA was used to calculate the main effects, the main effect index and the total sensitivity index for each parameter and these were compared to local sensitivity analysis results. The results suggest that the system performance is resistant to adverse changes in several parameters at once.

  4. Nonlinear Adaptive Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Dong, Zhe

    2013-04-01

    After the Fukushima nuclear accident, much more attention has to be drawn on the safety issues. The improvement of safety has already become the focus of the developing trend of the nuclear energy systems. Due to the inherent safety feature and the potential economic competitiveness, the modular high temperature gas-cooled reactor (MHTGR) has been seen as the central part of the next generation of nuclear plant (NGNP). Power-level control is one of the key techniques that guarantee the safe, stable and efficient operation for nuclear reactors. Since the MHTGR dynamics has the features of strong nonlinearity and uncertainty, in order to improve the operation performance, it is meaningful to develop the nonlinear adaptive power-level control law for the MHTGR. Based on using the natural dynamic features beneficial to system stabilization, a novel nonlinear adaptive power-level control is given for the MHTGR in this paper. It is theoretically proved that this newly-built controller does not only provide globally asymptotic closed-loop stability but is also adaptive to the system uncertainty. This control law is then applied to the power-level regulation of the pebble-bed MHTGR of the HTR-PM power plant. Numerical simulation results show the feasibility of this control law and the relationship between the performance and controller parameters.

  5. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  6. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  7. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  8. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  9. Some Experimental Investigations on Gas Turbine Cooling Performed with Infrared Thermography at Federico II

    Directory of Open Access Journals (Sweden)

    T. Astarita

    2015-01-01

    Full Text Available This paper reviews some experimental measurements of convective heat transfer coefficient distributions which are connected with the cooling of gas turbines, performed by the authors’ research group at the University of Naples Federico II with infrared thermography. Measurements concern impinging jets, cooling of rotating disks, and gas turbine blades, which are either stationary or rotating. The heated thin foil sensor, associated with the detection of surface temperature by means of infrared thermography, is exploited to accurately measure detailed convective heat transfer coefficient maps. The paper also intends to show how to correctly apply the infrared technique in a variety of gas turbines cooling problems.

  10. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  11. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  12. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  13. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wi S.; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2008-10-15

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO{sub 2}) driven Brayton cycle. According to their suggestion SCO{sub 2} is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider.

  14. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Lindemer, T.B.

    2002-09-05

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.

  15. A comparative study of the He and CO{sub 2} cycle for a small modular gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Seong Jun; Ahn, Yoon Han; Lee, Jeong Ik [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The gas-cooled nuclear reactor with closed Brayton cycle is considered as an attractive power conversion system because it can be compact and suitable system for reducing the total system size significantly while keeping the passive safety features. Helium and carbon dioxide (CO{sub 2}) are strong candidates as a coolant for the gas-cooled nuclear system. Helium Brayton cycle is commonly known that it can obtain very simple system arrangement with direct cycle and high thermal efficiency under high outlet temperature range due to its advantages such as less interaction with structure material, chemical stability and so on. However, supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle can be more suitable power conversion cycle with HTGR. The S-CO{sub 2} Brayton cycle has advantages over the helium Brayton cycle because it can achieve higher thermal efficiency at similar or even lower turbine inlet temperature (T. I. T) and can be more compact than a helium cycle. Both Brayton cycles can be a suitable power conversion system for a small modular gas-cooled reactor. Thus, for this study, preliminary design works of helium and the CO{sub 2} Brayton cycles for a 5MWth small modular gas-cooled reactor were carried out and evaluated while considering turbomachinery efficiency variation. Considering the size of a small modular nuclear system, the cycle configurations should be simple and compact. So, a simple recuperated Brayton cycle was chosen as candidate of the cycle layout for this study.

  16. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  17. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, January 1, 1979-March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-19

    This report presents the results of work performed from January 1, 1979 through March 31, 1979 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. Work covered in this report includes the activities associated with the creep-rupture testing of the test materials for the purpose of verifying the stresses selected for the screening creep test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment.

  18. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, October 1, 1978--December 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-09

    Results of work performed from October 1, 1978 through December 31, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program is presented. Objectives are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys, and selection of materials for future test facilities and more extensive qualification programs. The activities associated with the characterization of the materials for the screening test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are included. The status of the data management system is presented.

  19. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  20. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  1. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  2. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zheng Yanhua, E-mail: zhengyh@mail.tsinghua.edu.c [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Shi Lei; Wang Yan [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-10-15

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

  3. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  4. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  5. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    Science.gov (United States)

    Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

    2004-02-01

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  6. Operation and Control Simulation of a Modular High Temperature Gas Cooled Reactor Nuclear Power Plant

    Science.gov (United States)

    Li, Haipeng; Huang, Xiaojin; Zhang, Liangju

    2008-08-01

    Issues in the operation and control of the multi-modular nuclear power plant are complicated. The high temperature gas cooled reactor pebble-bed module (HTR-PM) plant with two-module will be built as a demonstration plant in China. To investigate the operation and control characteristics of the plant, a simplified dynamic model is developed and mathematically formulated based upon the fundamental conversation of mass, energy and momentum. The model is implemented in a personal computer to simulate the power increase process of the HTR-PM operation. The open loop operation with no controller is first simulated and the results show that the essential parameter steam temperature varies drastically with time, which is not allowable in the normal operation. According to the preliminary control strategy of the HTR-PM, a simple steam temperature controller is proposed. The controller is of Proportional-type with a time lag. The closed loop operation with a steam temperature controller is then implemented and the simulation results show that the steam temperature and also other parameters are all well controlled in the allowable range.

  7. Physically-Based Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Dong, Zhe

    2012-10-01

    Because of its strong inherent safety, the modular high temperature gas-cooled nuclear reactor (MHTGR) has been regarded as the central part of the next generation nuclear plants (NGNPs). Power-level control is one of the key techniques which provide safe, stable and efficient operation for the MHTGRs. The physically-based regulation theory is definitely a promising trend of modern control theory and provides a control design method that can suppress the unstable part of the system dynamics and remain the stable part. Usually, the control law designed by the physically-based control theory has a simple form and high performance. Stimulated by this, a novel nonlinear dynamic output feedback power-level control is established in this paper for the MHTGR based upon its own dynamic features. This newly-built control strategy guarantees the globally asymptotic stability and provides a satisfactory transient performance through properly adjusting the feedback gains. Simulation results not only verify the correctness of the theoretical results but also illustrate the high control performance.

  8. Utilization of heat of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ide, A. [Fuji Electric, Tokyo (Japan). Nuclear Power Promotion Dept.; Takenaka, Y. [Kawasaki Heavy Industries, Tokyo (Japan). Nuclear Systems Div.; Maeda, S. [Ube Industries, Yamaguchi (Japan). Machinery Dept.

    1996-07-01

    The demand for energy is increasing worldwide along with increases in population and rises in the standard of living. If the needed energy is supplied only by fossil fuels, environmental problems will impose limits on human activities. Recognizing that more than 60% of the energy consumed in Japan is non-electrical energy, FAPIG organized the HTR-HUC Working Group to study methods of using heat from high temperature gas-cooled reactors (HTR) to mitigate environmental and energy resource problems, and to contribute to the steady supply and effective use of energy. The authors chose three types of model plants to study: (1) a cogeneration plant which can be built with existing technology; (2) a coal gasification plant which can accelerate the clean use of coal and contribute to a stable supply of energy and the preservation of the environment; and (3) a hydrogen production plant whose hydrogen will release people from their dependence on fossil energy. For each of the above plants, a system outline and basic plan as well as costs, resultant social effects, management methods of the operating company and technical issues are studied.

  9. Cooling system for a nuclear reactor

    Science.gov (United States)

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  10. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  11. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Science.gov (United States)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  12. Helium-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  13. Application of optical fibers for optical diagnostics in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shikama, T.; Narui, M. [Oarai Branch, Institute for Materials Research, Tohoku University, Ibaraki-ken (Japan); Kakuta, T. [Tokai Research Establishment, JAERI, Ibaraki-ken (Japan); Ishihara, M.; Sagawa, T.; Arai, T. [Oarai Research Establishment, JAERI, Ibaraki-ken (Japan)

    1998-09-01

    Visibility of a core region of a high temperature gas cooled reactor (HTGR) is very poor in general with its solid graphite moderator. Realization of optical diagnostics will improve safety and maintenance of the HTGR considerably. The applicability of fused silica core optical fibers for optical diagnostics in a core of the High Temperature Testing Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been studied in the present research. Optical diagnostics are also expected to play crucial roles in advanced research planned in the HTTR. Optical transmission of the optical fibers was found not to degrade for several hundred hours at 1070K in air and helium environments in the wavelength range of 350-1800nm. In general. the optical fibers were found to be heat-resistant. To study radiation effects, the optical fibers were irradiated in Japan Materials Testing Reactor (JMTR). where the fast neutron(E>1MeV) flux was up to 1.5x10{sup 18}n/m{sup 2}s and the gamma-ray dose rate was up to about 5W/g for iron. The estimated fast neutron flux and the gamma-ray dose rate would be in the order of 10{sup 16}n/m{sup 2} and about 0.1W/g for iron, respectively in the HTTR. In general, optical transmission loss increased substantially with a small irradiation dose in the visible wave length range, although some developed fibers showed better radiation resistance. Good optical transmissivity was kept in the infrared region with absorption rate of less than a few dB/m. Radioluminescence and thermoluminescence from sapphire and silica could be observed with optical fibers under irradiation. Cherenkov radiation was observed in the wavelength range of 600-1800nm, whose intensity was temperature-independent. Black-body radiation was dominant in the wavelength longer than 1200nm at elevated temperatures. The results showed that the silica core optical fibers could be used as an image guide as well as monitors for radiation dosimetry and for monitoring core

  14. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  15. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  16. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D. A.; Hobbins, R. R.; Lowry, P.; Gougar, H.

    2013-11-01

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  17. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  18. Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Zuoyi; Scherer, W.

    1996-01-01

    This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H{sub 2}O concentration increase in the reactor core is smaller than 0.3 kg/(m{sup 3}s). The liquid water vaporization in the steam generator and H{sub 2}O transport from the steam generator to the reactor core reduce the ingress velocity of the H{sub 2}O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H{sub 2}O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [Deutsch] Dieser Bericht analysiert schwere Wassereinbruch-Stoerfaelle im 200 MW modularen Kugelhaufen-Hochtemperaturreaktor (HTR

  19. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  20. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  1. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  2. AIR COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  3. Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

    2003-01-01

    In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...

  4. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  5. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)

  6. Effect of fuel burnup and cross sections on modular HTGR (High-Temperature Gas-cooled Reactor) reactivity coefficients

    Science.gov (United States)

    Lefler, W.; Baxter, A.; Mathews, D.

    1987-12-01

    The temperature dependence of the reactivity coefficient in a prismatic block Modular High-Temperature Gas-Cooled Reactor (MHTGR) design is examined and found to be large and negative. Temperature coefficient results obtained with the ENDF/B-V data library were almost the same as results obtained with the earlier versions of the ENDF/B data library usually used at GA Technologies Inc., in spite of a significant eigenvalue increase with the ENDF/B-V data. The effects of fuel burnup and arbitrarily assumed cross section variations were examined and tabulated.

  7. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  8. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  9. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  10. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  11. Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

    1980-02-01

    Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed.

  12. Fissile compound - Inert matrix compatibility studies for the development of gas cooled fast reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Rapaud, O.; Rado, C

    2004-07-01

    Helium-Cooled High-Temperature Fast Reactors have a high potential for transmutation of minor actinides (Pu, Am, Cm... ). In this kind of reactor, the fuel temperature would be 1200 deg C in use and the inert matrix should retain the fission products in the fuel structure up to 1600 deg C. The fissile compound would be (U,Pu)C or (U,Pu)N owing to their high density, good thermal conductivity and refractory behavior. SiC, TiC, ZrC and TiN, ZrN would be the inert matrix surrounding (U,Pu)C or (U,Pu)N fissile compounds. This study is devoted to the chemical compatibility between UC or UN and inert matrix in the 1200 deg C - 2000 deg C temperature range. In order to achieve a limited number of specific experiments, thermodynamic calculations are realized using the thermodynamic data provided either by the Thermodata database or from the literature. (authors)

  13. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  14. Design study of gas cooled fast reactors using natural uranium as fuel cycle input employing radial shuffling strategy

    Science.gov (United States)

    Irka, Feriska Handayani; Su'ud, Zaki; Aryani, Menik; Aziz, Ferhat; Sekimoto, H.

    2012-06-01

    Design study of gas cooled fast reactors with natural uranium as fuel cycle input has been performed. The reactors utilizes UN-PUN as fuel, helium as coolant, and can be operated without refueling for 10 years in each batch. Reactor design optimization is performed to utilize natural uranium as fuel cycle input. This reactor subdivided into 10 regions with equal volume in radial directions. The natural uranium is initially put in region 1, and after one cycle of 10 years of burnup it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions. To achieve criticality requirement relatively high fuel volume fraction is applied. In this study several parametric survey were performed for several parameters such as fuel-to-coolant volume fraction ratio, core radius, and core height. After some optimization process we determine a standard core with a height and a diameter of 350 cm and 240 cm respectively, and the volume fraction for this design is 65% fuel, 10% cladding and 25% coolant. Calculation has been done by using SRAC-Citation system code and JENDL-3.2 library.

  15. Sensitivity study on depressurized LOFC accidents with failure of RCCS in a modular gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seyun [Nuclear Power Laboratory, Korea Electric Power Research Institute, Munji-ro 65, Yuseong, Daejeon 305-380 (Korea, Republic of); Tak, Nam-Il; Lim, Hong-Sik [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong, Deajeon 305-353 (Korea, Republic of); Ha, Sang-Jun [Nuclear Power Laboratory, Korea Electric Power Research Institute, Munji-ro 65, Yuseong, Daejeon 305-380 (Korea, Republic of)

    2010-05-15

    A modular gas-cooled reactor design with a thermal output of 600 MWt and a core exit temperature of 950 deg. C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multicomponent problems was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600 deg. C at about 50 h at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to

  16. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  17. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  18. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  19. Gas-cooled reactor commercialization study: introduction scenario and commercialization analyses for process heat applications. Final report, July 8, 1977--November 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    This report identifies and presents an introduction scenario which can lead to the operation of High Temperature Gas Cooled Reactor demonstration plants for combined process heat and electric power generation applications, and presents a commercialization analysis relevant to the organizational and management plans which could implement a development program.

  20. Evaluation of radiation heat transfer in porous medial: Application for a pebble bed modular reactor cooled by CO2 gas

    Directory of Open Access Journals (Sweden)

    Sidi-Ali Kamel

    2013-01-01

    Full Text Available This work analyses the contribution of radiation heat transfer in the cooling of a pebble bed modular reactor. The mathematical model, developed for a porous medium, is based on a set of equations applied to an annular geometry. Previous major works dealing with the subject have considered the forced convection mode and often did not take into account the radiation heat transfer. In this work, only free convection and radiation heat transfer are considered. This can occur during the removal of residual heat after shutdown or during an emergency situation. In order to derive the governing equations of radiation heat transfer, a steady-state in an isotropic and emissive porous medium (CO2 is considered. The obtained system of equations is written in a dimensionless form and then solved. In order to evaluate the effect of radiation heat transfer on the total heat removed, an analytical method for solving the system of equations is used. The results allow quantifying both radiation and free convection heat transfer. For the studied situation, they show that, in a pebble bed modular reactor, more than 70% of heat is removed by radiation heat transfer when CO2 is used as the coolant gas.

  1. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  2. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  3. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  4. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  5. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  6. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-12

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050/sup 0/C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850/sup 0/C on several experimental alloys are discussed.

  7. Sensitivity study on loss-of-forced-circulation accidents in a modular gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seyun; Ha, Sangjun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Tak, Namil; Lim, Hongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    The GCR has inherent safety characteristics with its large heat capacity and low power density of the core when compared with conventional light water reactors. Accordingly, a temperature change in a GCR core is very slow for a transient variation of the system temperature. One of the remarkable features of the modular GCR is a possibility to remove residual heat through the reactor vessel surface to the heat sink due to natural heat transfer processes. In the GCR system, graphite (IG-110) is used as a reflector and a core structure material. The uncertainties in the graphite material properties exist in its design and safety analysis processes. Sensitivity study on the major material properties which have an uncertainty in an LOFC accident condition and analyses on the thermal behavior of a reactor in the accident condition to assess the heat transfer characteristics are carried out.

  8. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable

  9. Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

    Science.gov (United States)

    Gorensek, Maximilian B

    2015-02-17

    A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

  10. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Lee Nelson

    2009-10-01

    This report is a preliminary comparison of conventional and potential HTGR-integrated processesa in several common industrial areas: ? Producing electricity via a traditional power cycle ? Producing hydrogen ? Producing ammonia and ammonia-derived products, such as fertilizer ? Producing gasoline and diesel from natural gas or coal ? Producing substitute natural gas from coal, and ? Steam-assisted gravity drainage (extracting oil from tar sands).

  11. Methods of assessing the effects of interface oxide growth in Magnox and advanced gas-cooled reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    McLauchlin, I.R.; Wooton, M.R.; Morgan, J.D.; Watson, L.H.

    1986-10-01

    Growth of oxide at interfaces between structural steel components in CO/sub 2/-cooled reactors can deform fastenings such as bolts and welds. The mechanical response of joint members to oxide growth is discussed, and methods of assessment are outlined which contribute to procedures for ensuring continued structural integrity.

  12. Design study of a modular gas-cooled, closed-brayton cycle reactor for marine use

    Science.gov (United States)

    Lantz, Richard D.

    1989-06-01

    A conceptual design of a direct Brayton cycle marine power plant is presented. The design is a modification of the commercial MGR-GT, as proposed by James Staudt, sized to produce 40,000 shaft horsepower (SHP) and 5 MW of ship service electrical power. The requirements of a shipboard power plant are discussed and the design changes that must be made to the components of a commercial power plant in order to fit them into the demanding environment of a ship at sea are detailed. The final design consists of an 80-MWth passively safe pebble bed reactor with an outlet temperature of 850.

  13. Output feedback dissipation control for the power-level of modular high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Z. [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China)

    2011-07-01

    Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR) is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis. (author)

  14. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2011-11-01

    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  15. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  16. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  17. A COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Creutz, E.C.

    1960-03-15

    A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

  18. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  19. Inquiry into disintegration control of irradiated low enrichment uranium for high temperature gas cooled reactors

    Science.gov (United States)

    Reitsamer, G.; Stolba, G.; Falta, G.; Strigl, A.; Zeger, J.; Maly, V.

    1984-07-01

    The PyC-coatings of irradiated high temperature reactor (HTR) fuel particles from AVR fuel elements were burnt off by air and oxygen at 1120K (comparable to the current HTR head-end). Experiments on the solubility of this low enrichment uranium fuel prove that 99.93% of the uranium and 99.84% of the plutonium can be dissolved by 7n HNO3. After additional treatment with 7n HNO3/0.01 n NaF, only 0.01% of the original amount of uranium and 0.01% of the original amount of plutonium remain undissolved. Neither the insoluble residues nor the very small amounts of solids formed on standing (before and after concentrating the solution up to 200 g U/1 and acidity of 3 n HNO3) show any enrichment of plutonium compared with the nitric acid solution. Results indicate that LWR-PUREX-technology can be used for reprocessing HTR-LEU-fuel.

  20. FACTORS INFLUENCING HUMAN RELIABILITY OF HIGH TEMPERATURE GAS COOLED REACTOR OPERATION

    Directory of Open Access Journals (Sweden)

    Sigit Santoso

    2016-10-01

    ABSTRAK Peran dan tindakan operator pada reaktor berpendingin gas akan berbeda dengan peran operator pada operasi tipe reaktor lain. Analisis unjuk kerja operator dan faktor yang berpengaruh dapat dilakukan secara komprehensif melalui analisis keandalan manusia(HRA. Melalui HRA dampak dari kesalahan manusia pada sistem maupun cara untuk mengurangi dampak dan frekuensi kesalahan dapat diketahui. Makalah membahas faktor yang berpengaruh pada tindakan operator, yaitu pada kejadian kecelakaan pendingin reaktor gas bersuhu tinggi-HTGR. Analisis untuk kualifikasi faktor pembentuk kinerja(PSF dilakukan berdasarkan kurva keandalan fungsi waktu, dan metode keandalan manusia yang dikembangkan berdasar pada aspek kognitif yaitu Cognitive Reliability and Error Analysis Method (CREAM. Hasil analisis berdasar kurva keandalan fungsi waktu menunjukkan komponen waktu berkontribusi positif pada peningkatan keandalan operator (PSF<1 pada kondisi semua fitur keselamatan berfungsi sesuai rancangan. Sedangkan pada metoda analisis dengan pendekatan kognitif CREAM diketahui selain faktor ketersediaan waktu, faktor pelatihan dan rancangan HMI juga berkontribusi meningkatkan keandalan operator. Faktor pembentuk kinerja keseluruhan diketahui sebesar 0,25 dengan faktor kontribusi positif dominan atau berpengaruh pada penurunan kesalahan manusia adalah ketersediaan waktu (PSF=0,01, dan faktor kontribusi negatif dominan adalah prosedur dan siklus kerja (PSF=5. Nilai PSF tersebut sebagai faktor pengali dalam perhitungan probabilitas kesalahan manusia. Analisis faktor pembentuk kinerja perlu dikembangkan pada skenario kejadian lain untuk selanjutnya digunakan untuk perhitungan dan analisis keandalan manusia yang komprehensif dan perancangan sistem interaksi manusia mesin di ruang kendali. Kata kunci: PSF, HTGR, operator, ruang kendali, keandalan manusia

  1. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  2. High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Cole, T.E.; Sanders, J.P.; Kasten, P.R.

    1977-07-01

    Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.

  3. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  4. Integration of High Temperature Gas-cooled Reactor Technology with Oil Sands Processes

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation of siting an HTGR plant in a remote area supplying steam, electricity and high temperature gas for recovery and upgrading of unconventional crude oil from oil sands. The area selected for this evaluation is the Alberta Canada oil sands. This is a very fertile and active area for bitumen recovery and upgrading with significant quantities piped to refineries in Canada and the U.S Additionally data on the energy consumption and other factors that are required to complete the evaluation of HTGR application is readily available in the public domain. There is also interest by the Alberta oil sands producers (OSP) in identifying alternative energy sources for their operations. It should be noted, however, that the results of this evaluation could be applied to any similar oil sands area.

  5. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  6. Burnable poison calculations for Mk.III gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Gubbins, M.E.

    1971-02-15

    A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.

  7. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    This report documents developments and results in the frame of the project RS1191 ''Development of computational methods for the safety assessment of gas-cooled high temperature and supercritical light-water reactors''. The report is structured according to the five work packages: 1. Reactor physics modeling of gas-cooled high temperature reactors; 2. Coupling of reactor physics and 3-D thermal hydraulics for the core barrel; 3. Extension of ATHLET models for application to supercritical reactors (HPLWR); 4. Further development of ATHLET for application to HTR; 5. Further development and validation of ANSYS CFX for application to alternative reactor concepts. Chapter 4 describes the extensions made in TORT-TD related to the simulation of pebble-bed HTR, e.g. spectral zone buckling, Iodine-Xenon dynamics, nuclear decay heat calculation and extension of the cross section interpolation algorithms to higher dimensions. For fast running scoping calculations, a time-dependent 3-D diffusion solver has been implemented in TORT-TD. For the PBMR-268 and PBMR-400 as well as for the HTR-10 reactor, appropriate TORT-TD models have been developed. Few-group nuclear cross sections have been generated using the spectral codes MICROX- 2 and DRAGON4. For verification and validation of nuclear cross sections and deterministic reactor models, MCNP models of reactor core and control rod of the HTR-10 have been developed. Comparisons with experimental data have been performed for the HTR-10 first criticality and control rod worth. The development of the coupled 3-D neutron kinetics and thermal hydraulics code system TORT-TD/ATTICA3D is documented in chapter 5. Similar to the couplings with ATHLET and COBRA-TF, the ''internal'' coupling approach has been implemented. Regarding the review of experiments and benchmarks relevant to HTR for validation of the coupled code system, the PBMR-400 benchmarks and the HTR-10 test reactor have been selected

  8. Particle image velocimetry measurements in a representative gas-cooled prismatic reactor core model for the estimation of bypass flow

    Science.gov (United States)

    Conder, Thomas E.

    Core bypass flow is considered one of the largest contributors to uncertainty in fuel temperature within the Modular High Temperature Gas-cooled Reactor (MHTGR). It refers to the coolant that navigates through the interstitial regions between the graphite fuel blocks instead of traveling through the designated coolant channels. These flows are of concern because they reduce the desired flow rates in the coolant channels, and thereby have significant influence on the maximum fuel element and coolant exit temperatures. Thus, accurate prediction of the bypass flow is important because it directly impacts core temperature, influencing the life and efficiency of the reactor. An experiment was conducted at Idaho National Laboratory to quantify the flow in the coolant channels in relation to the interstitial gaps between fuel blocks in a representative MHTGR core. Particle Image Velocimetry (PIV) was used to measure the flow fields within a simplified model, which comprised of a stacked junction of six partial fuel blocks with nine coolant tubes, separated by a 6mm gap width. The model had three sections: The upper plenum, upper block, and lower block. Model components were fabricated from clear, fused quartz where optical access was needed for the PIV measurements. Measurements were taken in three streamwise locations: in the upper plenum and in the midsection of the large and small fuel blocks. A laser light sheet was oriented parallel to the flow, while velocity fields were measured at millimeter intervals across the width of the model, totaling 3,276 PIV measurement locations. Inlet conditions were varied to incorporate laminar, transition, and turbulent flows in the coolant channels---all which produced laminar flow in the gap and non-uniform, turbulent flow in the upper plenum. The images were analyzed to create vector maps, and the data was exported for processing and compilation. The bypass flow was estimated by calculating the flow rates through the coolant

  9. Gas turbine cooling system

    Science.gov (United States)

    Bancalari, Eduardo E.

    2001-01-01

    A gas turbine engine (10) having a closed-loop cooling circuit (39) for transferring heat from the hot turbine section (16) to the compressed air (24) produced by the compressor section (12). The closed-loop cooling system (39) includes a heat exchanger (40) disposed in the flow path of the compressed air (24) between the outlet of the compressor section (12) and the inlet of the combustor (14). A cooling fluid (50) may be driven by a pump (52) located outside of the engine casing (53) or a pump (54) mounted on the rotor shaft (17). The cooling circuit (39) may include an orifice (60) for causing the cooling fluid (50) to change from a liquid state to a gaseous state, thereby increasing the heat transfer capacity of the cooling circuit (39).

  10. A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A., E-mail: aaron@epiney.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Service d' Etudes des Systemes Innovants, F-13108 Saint Paul Lez Durance (France); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer An analytical model predicting Brayton cycle off-design steady states, is developed. Black-Right-Pointing-Pointer The model is used to design an autonomous decay heat removal system for the GFR. Black-Right-Pointing-Pointer Predictions of the analytical model are verified using CATHARE. Black-Right-Pointing-Pointer CATHARE code is used to simulate a set of GFR safety depressurization transients using this device. Black-Right-Pointing-Pointer Convenient turbo-machine designs exist for the targeted autonomous decay heat removal for a wide pressure range. - Abstract: This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping model', is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the

  11. Liquid metal cooled nuclear reactor plant system

    Science.gov (United States)

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  12. Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Taiwo, T. A.; Cahalan, J. E.; Finck, P. J.

    2001-05-08

    An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable absorber for both suppressing the initial excess

  13. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

    The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.

  14. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1978--March 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-06-26

    The activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials received, and preliminary information from the materials characterization tests performed by GE are reported. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The final recommended impurity levels for the screening phase helium are presented and the rational behind this gas chemistry is discussed. The status of the data management system is presented.

  15. State of Art Report for the Bypass and Cross Flows in Prismatic Modular High-Temperature Gas-Cooled Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Kim, Min Hwan

    2010-01-15

    The horizontal and vertical gaps between the adjacent fuel blocks occurs significantly in the prismatic modular high-temperature gas-cooled reactor core due to the thermal expansion, irradiation expansion/shrinking, and fuel block column bowing by pressure and gravity during the plant operation, in addition to the initial manufacture/design tolerance of fuel/reflector blocks. The coolant leakage of helium gas occurs through the gaps. These bypass and cross flows highly impact on the effective core cooling flow rate. This report describes the technical state of the bypass and cross flow study, based on ten reference papers, reviewing and summarizing four classified contents in the viewpoints of 'The reactor core fluctuation data for the first identification of the importance of the bypass and cross flow', 'The cross flow test and evaluation data', 'The flow test and evaluation data of the seal mechanism to prevent the leakage flow', and 'The reactor core thermal-fluid analysis and evaluation dat000.

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-18

    This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

  17. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  18. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  19. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  20. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  1. Experimental investigation of gas turbine airfoil aerodynamic performance without and with film cooling in an annular sector cascade

    Energy Technology Data Exchange (ETDEWEB)

    Wiers, S.H.

    2002-02-01

    The steady growing of industrialization, the densification of the anthroposphere, the increasing concern over the effects of gas turbine cruise emissions on the atmosphere threaten the growth of air transportation, and the perception about the possible climatic impact of CO{sub 2} emissions causes a public distinctive sense of responsibility. The conventional energy production techniques, which are based on fossil fuel, will keep its central importance within the global energy production. Forecasts about the increasing air transportation give duplication in the next 10-15 years. The optimization of the specific fuel consumption is necessary to decrease the running costs and the pollution emissions in the atmosphere, which makes an increased process efficiency of stationary turbines as well as of jet engines essential. This leads to the necessity of an increased thermodynamic efficiency of the overall process and the optimization of the aerodynamic components. Due to the necessity of more detailed three-dimensional data on the behavior of film cooled blades an annular sector cascade turbine test facility has gone into service. The annular sector cascade facility is a relative cost efficient solution compared to a full annular facility to investigate three-dimensional effects on a non cooled and cooled turbine blade. The aerodynamic investigations on the annular sector cascade facility are part of a broad perspective where experimental data from a hot annular sector cascade facility and the cold annular sector facility are used to verify, calibrate and understand the physics for both internal and external calculation methods for flow and heat transfer prediction. The objective of the present study is the design and validation of a cold flow annular sector cascade facility, which meets the flow conditions in a modem turbine as close as possible, with emphasis on achieving periodic flow conditions. The first part of this study gives the necessary background on this

  2. A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

    Energy Technology Data Exchange (ETDEWEB)

    Minatsuki, Isao, E-mail: isao_minatsuki@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Otani, Tomomi; Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan)

    2014-05-01

    A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies.

  3. The role of clusters in gas-solids reactors. An experimental study.

    NARCIS (Netherlands)

    Venderbosch, R.H.

    1998-01-01

    This PhD-work is meant to determine the contact efficiency experimentally for fluidization of fine particles over a wide range of superficial gas velocities (dp<200 mm and 0.1

  4. PARTICLE IMAGE VELOCIMETRY MEASUREMENTS IN A REPRESENTATIVE GAS-COOLED PRISMATIC REACTOR CORE MODEL: FLOW IN THE COOLANT CHANNELS AND INTERSTITIAL BYPASS GAPS

    Energy Technology Data Exchange (ETDEWEB)

    Thomas E. Conder; Richard Skifton; Ralph Budwig

    2012-11-01

    Core bypass flow is one of the key issues with the prismatic Gas Turbine-Modular Helium Reactor, and it refers to the coolant that navigates through the interstitial, non-cooling passages between the graphite fuel blocks instead of traveling through the designated coolant channels. To determine the bypass flow, a double scale representative model was manufactured and installed in the Matched Index-of-Refraction flow facility; after which, stereo Particle Image Velocimetry (PIV) was employed to measure the flow field within. PIV images were analyzed to produce vector maps, and flow rates were calculated by numerically integrating over the velocity field. It was found that the bypass flow varied between 6.9-15.8% for channel Reynolds numbers of 1,746 and 4,618. The results were compared to computational fluid dynamic (CFD) pre-test simulations. When compared to these pretest calculations, the CFD analysis appeared to under predict the flow through the gap.

  5. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    OpenAIRE

    Cisneros, Anselmo Tomas

    2013-01-01

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids - flibe (33%7Li2F-67%BeF) - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate i...

  6. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    OpenAIRE

    Udiyani, P.M; Sri Kuntjoro

    2017-01-01

    Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimati...

  7. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

  8. BLAST: a digital computer program for the dynamic simulation of the high temperature gas cooled reactor reheater-steam generator module

    Energy Technology Data Exchange (ETDEWEB)

    Hedrick, R.A.; Cleveland, J.C.

    1976-06-24

    BLAST simulates the high temperature gas cooled reactor reheater-steam generator module with a multi-node, fixed boundary, homogenous flow model. The time dependent conservation of energy, mass, and momentum equations are solved by an implicit integration technique. The code contains equation of state formulations for both helium and water as well as heat transfer and friction factor correlations. Normal operational transients and more severe transients such as those resulting in low and/or reverse flow can be simulated. The code calculates helium and water temperature, pressure, flow rate, and tube bulk and wall temperatures at various points within the reheater-steam generator module during the transients. BLAST predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in BLAST. BLAST is written in FORTRAN IV for the IBM 360/91 computer at the Oak Ridge National Laboratory.

  9. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  10. Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

    2006-05-15

    For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

  11. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Quarterly progress report, April 1, 1977--June 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The objectives of this program are to evaluate candidate alloys of Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle helium Turbine (DCHT) applications, in terms of the effects of simulated reactor primary coolant (impure helium), high temperatures, and long time exposures on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes completion of alloy selection for the screening tests. The alloys selected for potential VHTR Nuclear Process Heat (NPH) applications and for potential DCHT applications are listed. The present status on the simulated reactor helium loop design and construction and the design and construction progress on the testing and analysis facilities and equipment are discussed.

  12. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1, 1977--September 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-11-14

    Work covered includes an updated listing of the alloys selected for the screening tests, plus complete test specimen matrices for the screening program. The present design and construction status of the simulated reactor helium loops and testing and analysis facilities and equipment are discussed. Also covered are the loading matrices for the screening creep tests.

  13. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  14. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    Energy Technology Data Exchange (ETDEWEB)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  15. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Roguin, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

  16. Method and apparatus for enhancing reactor air-cooling system performance

    Science.gov (United States)

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  17. A Secondary Flow Effect on the Heat and Mass Transfer Processes in the Finned Rod Bundles of Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Dunaitsev

    2017-01-01

    Full Text Available In nuclear power engineering a need to justify an operability of products and their components is of great importance. In high-temperature gas reactors, the critical element affecting the facility reliability is the fuel rod cladding, which in turn leads to the need to gain knowledge in the field of gas dynamics and heat transfer in the reactor core and to increase the detail of the calculation results. For the time being, calculations of reactor core are performed using the proven techniques of per-channel calculations, which show good representativeness and count rate. However, these techniques require additional experimental studies to describe correctly the inter-channel exchange, which, being taken into account, largely affects the pattern of the temperature fields in the region under consideration. Increasingly more relevant and demandable are numerical simulation methods of fluid and gas dynamics, as well as of heat exchange, which consist in the direct solution of the system of differential equations of mass balance, kinetic moment, and energy. Calculation of reactor cores or rod bundles according these techniques does not require additional experimental studies and allows us to obtain the local distributions of flow characteristics in the bundle and the flow characteristics that are hard to measure in the physical experiment.The article shows the calculation results and their analysis for an infinite rod lattice of the reactor core. The results were obtained by the technique of modelling one rod of a regular lattice using the periodic boundary conditions, followed by translating the results to the neighbouring rods. In channels of complex shape, there are secondary flows caused by changes in the channel geometry along the flow and directed across the main front of the flow. These secondary flows in the reactor cores with rods spaced by the winding wire lead to a redistribution of the coolant along the channel section, which in turn

  18. High temperature gas cooled reactor applications and future prospects. Proceedings of an IAEA Technical Committee Meeting held in Petten, the Netherlands, 10-12 November 1997

    Energy Technology Data Exchange (ETDEWEB)

    Haverkate, B.R.W. [ed.

    1998-09-01

    From 10-12 November, 1997, the Netherlands Energy Research Foundation (ECN) in Petten, Netherlands hosted a Technical Committee Meeting (TCM) on High Temperature Gas Cooled Reactors (HTGR). This meeting has been organised by the International Atomic Energy Agency (IAEA) and was entitled: `HTGR Applications and Future Prospects`. During the meeting a review of the status of national programmes, including the design and construction of HTGR plants, status of R and D programmes and related activities in support of the advancement and applications of the HTGR have been reported. 21 papers were presented in three sessions, respectively: nine papers in the first session Status of GCR Programmes, seven papers in the session HTGR Applications and five papers in the last session HTGR Development Activities. The meeting has been attended by approximately fifty participants from nine countries all over the world. The Nuclear Energy Agency (NEA) of the OECD and the European Commission have also attended this TCM. The IAEA TCM was followed, from 12-14 November, 1997, by an OECD/NEA workshop on High Temperature Engineering Research Facilities and Experiments to complement and support the IAEA activities in the HTGR field. The proceedings of this workshop have also been published by ECN as report ECN-R-98-005. 15 refs.

  19. A Critical Review of the Recent Improvements in Minimizing Nuclear Waste by Innovative Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    E. Bomboni

    2008-01-01

    Full Text Available This paper presents a critical review of the recent improvements in minimizing nuclear waste in terms of quantities, long-term activities, and radiotoxicities by innovative GCRs, with particular emphasis to the results obtained at the University of Pisa. Regarding these last items, in the frame of some EU projects (GCFR, PUMA, and RAPHAEL, we analyzed symbiotic fuel cycles coupling current LWRs with HTRs, finally closing the cycle by GCFRs. Particularly, we analyzed fertile-free and Pu-Th-based fuel in HTR: we improved plutonium exploitation also by optimizing Pu/Th ratios in the fuel loaded in an HTR. Then, we chose GCFRs to burn residual MA. We have started the calculations on simplified models, but we ended them using more “realistic” models of the reactors. In addition, we have added the GCFR multiple recycling option using keff calculations for all the reactors. As a conclusion, we can state that, coupling HTR with GCFR, the geological disposal issues concerning high-level radiotoxicity of MA can be considerably reduced.

  20. Small Liquid Metal Cooled Reactor Safety Study

    Energy Technology Data Exchange (ETDEWEB)

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  1. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

    OpenAIRE

    Girardin, Gaëtan

    2009-01-01

    During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. It is in this context that the Generation IV international forum (GIF) launched the initiative, in 2000, to collaborate on the research and development (R&D) efforts needed for the next generation, i.e. Generation I...

  2. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  3. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1, 1982-June 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    1982-10-01

    Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is described; this includes screening creep results and metallographic analyses for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/ and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in controlled-purity helium and air in the intensive screening test program is discussed. The results of metallographic studies of corrosion pins exposed at 750/sup 0/C for 6000 hours in controlled-purity helium (solid solution strengthened alloys, centrifugally cast alloys and an iron-base oxide dispersion strengthened alloy) are presented and discussed. The results of metallographic studies on the same materials after 10,000 hour exposure in controlled-purity helium at 850/sup 0/C are also presented.

  4. IDAHO NATIONAL LABORATORY PROGRAM TO OBTAIN BENCHMARK DATA ON THE FLOW PHENOMENA IN A SCALED MODEL OF A PRISMATIC GAS-COOLED REACTOR LOWER PLENUM FOR THE VALIDATION OF CFD CODES

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a typical prismatic gas-cooled (GCR) reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A detailed description of the model, scaling, the experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that are presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic GCR design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal undeveloped, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet flow is also presented.

  5. Study of Natural Convection Passive Cooling System for Nuclear Reactors

    Science.gov (United States)

    Abdillah, Habibi; Saputra, Geby; Novitrian; Permana, Sidik

    2017-07-01

    Fukushima nuclear reactor accident occurred due to the reactor cooling pumps and followed by all emergencies cooling systems could not work. Therefore, the system which has a passive safety system that rely on natural laws such as natural convection passive cooling system. In natural convection, the cooling material can flow due to the different density of the material due to the temperature difference. To analyze such investigation, a simple apparatus was set up and explains the study of natural convection in a vertical closed-loop system. It was set up that, in the closed loop, there is a heater at the bottom which is representing heat source system from the reactor core and cooler at the top which is showing the cooling system performance in room temperature to make a temperature difference for convection process. The study aims to find some loop configurations and some natural convection performances that can produce an optimum flow of cooling process. The study was done and focused on experimental approach and simulation. The obtained results are showing and analyzing in temperature profile data and the speed of coolant flow at some point on the closed-loop system.

  6. Analysis of the gas-diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Z.; Gerwin, H.; Scherer, W.

    1993-10-01

    A one-dimensional coupled diffusion-convection model has been established and it is shown, that experiments performed at JAERI have been recalculated successfully, thus validating the new model. Applying this model to the NACOK facility, now under construction at the Institute for Safety Research and Reactor Technology (ISR) of the KFA Juelich, a delay time until the onset of natural circulation was found to be 14 hours. For the 200 MW HTR-MODULE designed by the SIEMENS company about 39 hours are predicted. It was found that pressure disturbances, e.g. caused by environmental noise could significantly shorten the time delay until the onset of natural circulation. The onset time delay remains constant below 40 dB noise, reaches half of its value at 80 dB and tends to zero at 120 dB. Thus, the relevance of the delay effect with respect to safety related questions is reduced. (orig./HP) [Deutsch] Es wurde ein eindimensionales gekoppeltes Diffusions-Konvektions-Rechenmodell erstellt und zunaechst gezeigt, dass zur Validierung dieses Modells Experimente, welche am JAERI durchgefuehrt worden sind, mit gutem Erfolg nachgerechnet werden konnten. Die Anwendung des Modells auf das im ISR des KFA im Aufbau befindliche Experiment NACOK ergibt eine Verzoegerungszeit bis zum Einsatz der Naturkonvektion von 14 Stunden. Fuer den 200 MW HTR-MODUL der Firma SIEMENS werden etwa 39 Stunden vorhergesagt. Druckschwankungen, etwa hervorgerufen durch Umgebungslaerm, koennen diese Karenzzeit jedoch erheblich verkuerzen. Unterhalb von 40 dB Laerm bleibt die Verzoegerungszeit konstant, bei 80 dB halbiert sie sich und geht bei 120 dB gegen Null. Damit wird die sicherheitstechnische Nutzbarkeit dieses Verzoegerungseffektes eingeschraenkt. (orig./HP)

  7. An integrated systems calculation of a steam generator tube rupture in a modular prismatic HTGR (high-temperature gas-cooled reactor) conceptual design using ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer)

    Energy Technology Data Exchange (ETDEWEB)

    Beelman, R.J. (Idaho National Engineering Laboratory, Idaho Falls (USA))

    1989-11-01

    The capability to perform integrated systems calculations of modular high-temperature gas-cooled reactor (MHTGR) transients has been developed at the Idaho National Engineering Laboratory (INEL) using the Advanced Thermal-Hydraulic Energy Network Analyzer (ATHENA) computer code. A scoping calculation of a steam generator tube rupture (SGTR) water ingress event in a prismatic 2 {times} 350-MW(thermal) MHTGR conceptual design has been completed at INEL using ATHENA. The proposed MHTGR design incorporates dual, graphite-moderated, helium-cooled, 350-MW(thermal), annular prismatic core concept reactor plants, each configured with an individual helical once-through steam generator steaming a common 280-MW(electric) turbine generator set.

  8. Research of the influence of intensification of heat transfer on distribution of temperature in the active core of the gas cooled nuclear reactor of the «GT-MHR» project

    Science.gov (United States)

    Kuzevanov, V. S.; Podgorny, S. K.

    2017-11-01

    The maximum wall temperature of a cooling channel of a nuclear reactor is one of the factors that affects directly of the safety and reliability of the nuclear reactor. In this paper suggested an equation, which allows calculating the maximum wall temperature of the cooling channel of the nuclear reactor with heat transfer enhancer installed, without enormous calculations.

  9. Proposals of new basic concepts on safety and radioactive waste and of new High Temperature Gas-cooled Reactor based on these basic concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro, E-mail: ogawa.masuro@jaea.go.jp

    2016-11-15

    Highlights: • The author proposed new basic concepts on safety and radioactive waste. • A principle of ‘continue confining’ to realize the basic concept on safety is also proposed. • It is indicated that only a HTGR can attain the conditions required from the principle. • Technologies to realize the basic concept on radioactive waste are also discussed. • A New HTGR system based on the new basic concepts is proposed. - Abstract: A new basic concept on safety of ‘Not causing any serious catastrophe by any means’ and a new basic concept on radioactive waste of ‘Not returning any waste that possibly affects the environment’ are proposed in the present study, aiming at nuclear power plants which everybody can accept, in consideration of the serious catastrophe that happened at Fukushima Japan in 2011. These new basic concepts can be found to be valid in comparison with basic concepts on safety and waste in other industries. The principle to realize the new basic concept on safety is, as known well as the inherent safety, to use physical phenomena such as Doppler Effect and so on which never fail to work even if all equipment and facilities for safety lose their functions. In the present study, physical phenomena are used to ‘continue confining’, rather than ‘confine’, because the consequence of emission of radioactive substances to the environment cannot be mitigated. To ‘continue confining’ is meant to apply natural correction to fulfill inherent safety function. Fission products must be detoxified to realize the new basic concept on radioactive waste, aiming at the final processing and disposal of radioactive wastes as same as that in the other wastes such as PCB, together with much efforts not to produce radioactive wastes and to reduce their volume nevertheless if they are emitted. Technology development on the detoxification is one of the most important subjects. A new High Temperature Gas-cooled Reactor, namely the New HTGR

  10. An Experimental and Numerical Investigation of Endwall Aerodynamics and Heat Transfer in a Gas Turbine Nozzle Guide Vane with Slot Film Cooling

    Science.gov (United States)

    Alqefl, Mahmood Hasan

    In many regions of the high-pressure gas turbine, film cooling flows are used to protect the turbine components from the combustor exit hot gases. Endwalls are challenging to cool because of the complex system of secondary flows that disturb surface film coolant coverage. The secondary flow vortices wash the film coolant from the surface into the mainstream significantly decreasing cooling effectiveness. In addition to being effected by secondary flow structures, film cooling flow can also affect these structures by virtue of their momentum exchange. In addition, many studies in the literature have shown that endwall contouring affects the strength of passage secondary flows. Therefore, to develop better endwall cooling schemes, a good understanding of passage aerodynamics and heat transfer as affected by interactions of film cooling flows with secondary flows is required. This experimental and computational study presents results from a linear, stationary, two-passage cascade representing the first stage nozzle guide vane of a high-pressure gas turbine with an axisymmetrically contoured endwall. The sources of film cooling flows are upstream combustor liner coolant and endwall slot film coolant injected immediately upstream of the cascade passage inlet. The operating conditions simulate combustor exit flow features, with a high Reynolds number of 390,000 and approach flow turbulence intensity of 11% with an integral length scale of 21% of the chord length. Measurements are performed with varying slot film cooling mass flow to mainstream flow rate ratios (MFR). Aerodynamic effects are documented with five-hole probe measurements at the exit plane. Heat transfer is documented through recovery temperature measurements with a thermocouple. General secondary flow features are observed. Total pressure loss measurements show that varying the slot film cooling MFR has some effects on passage loss. Velocity vectors and vorticity distributions show a very thin, yet intense

  11. Liquid metal cooled reactors for space power applications

    Science.gov (United States)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  12. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  13. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  14. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  15. Experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors; Experimentelle und numerische Untersuchung von Gas/Liquid-Phasengrenzflaechen als Referenzwert fuer die hydrostatische Fuellstandsmessung in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, Stephan

    2013-12-17

    The experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors is considered as relevant for reactor safety research. The experiments allow a quantification of the transition processes in hydrostatic level measurement devices that were up to now only assessed by phenomenological descriptions. Experimental studies covered the topology and stability of water/vapor phase boundaries and the numerical description using CFD codes, including modeling of the surface topology and modeling of the heat and mass transport.

  16. Impingement jet cooling in gas turbines

    CERN Document Server

    Amano, R S

    2014-01-01

    Due to the requirement for enhanced cooling technologies on modern gas turbine engines, advanced research and development has had to take place in field of thermal engineering. Impingement jet cooling is one of the most effective in terms of cooling, manufacturability and cost. This is the first to book to focus on impingement cooling alone.

  17. Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

    2009-10-15

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  18. Modelling and experimental validation of the hot-gas defrost process of an air-cooled evaporator

    Energy Technology Data Exchange (ETDEWEB)

    Dopazo, J. Alberto; Fernandez-Seara, Jose; Uhia, Francisco J.; Diz, Ruben [Area de Maquinas y Motores Termicos, E.T.S. de Ingenieros Industriales, University of Vigo, Campus Lagoas-Marcosende No 9, 36310 Vigo, Pontevedra (Spain)

    2010-06-15

    A detailed transient simulation model has been developed to predict and evaluate the performance of the hot-gas defrost process of an air-coil evaporator. In the model, the defrost process is subdivided into six stages: preheating, tube frost melting start, fin frost melting start, air presence, tube-fin water film and dry-heating. In each stage, the control volume is subdivided into systems represented by a single node, which has the representative properties of the system. A finite difference approach was used to solve the model equations. The results include the time required to defrost, the distribution of the energy during defrost process, the instantaneous refrigerant properties and the instantaneous fin and tube temperature distribution. The results are compared with experimental data obtained in a local storage facility under actual operating conditions and also using data available in the literature. The model results substantially agree with the experimental data in both cases. (author)

  19. Gas pollutant cleaning by a membrane reactor

    Directory of Open Access Journals (Sweden)

    Kaldis Sotiris

    2006-01-01

    Full Text Available An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al2O3 catalyst was prepared by the dry and wet impregnation method and characterized by the inductively coupled plasma method, scanning electron microscopy, X-ray diffraction, and N2 adsorption before and after activation. Commercially available a-Al2O3 membranes were also characterized and the permeabilities and permselectivities of H2, N2, and CO2 were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. .

  20. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  1. Monitoring system for a liquid-cooled nuclear fission reactor

    Science.gov (United States)

    DeVolpi, Alexander

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  2. Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

    2016-04-15

    Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

  3. Reactor core and plant design concepts of the Canadian supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Bailey, J.; Rhodes, D.; Guzonas, D.; Hamilton, H.; Haque, Z.; Pencer, J.; Sartipi, A. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada is developing a 1200 MWe supercritical water-cooled reactor (SCWR), which has evolved from the well-established pressure-tube type CANDU{sup 1} reactor. This SCWR reactor concept, which is often referred to as the Canadian SCWR, uses supercritical water as a coolant, has a low-pressure heavy water moderator and a direct cycle for power production. The reactor concept incorporates advanced safety features, such as passive emergency core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce the core damage frequency beyond existing nuclear reactors. This paper presents a description of the Canadian SCWR core design concept, the integration of in-core and out-of-core components and the mechanical plant design concept. Supporting systems for reactor safety, reactor control and moderator cooling are also described. (author)

  4. In-line monitoring of effluents from high-temperature gas-cooled reactor fuel particle preparation processes by mass spectrometry. [UO/sub 2/; UC/sub 2/

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.A.; Constanzo, D.A.; Stinton, D.P.; Carpenter, J.A. Jr.; Rainey, W.T.; Canada, D.C.; Carter, J.A.

    1977-06-01

    The carbonization, conversion, and coating processes in the manufacture of high-temperature gas-cooled reactor fuel particles have been studied with the use of a time-of-flight mass spectrometer. Noncondensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO/sub 2/ to UC/sub 2/, buffer and low-temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures.

  5. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

    Energy Technology Data Exchange (ETDEWEB)

    El Ganaoui, K

    2006-09-15

    In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

  6. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  7. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  8. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    OpenAIRE

    Shirvan, Koroush; Kazimi, Mujid

    2016-01-01

    A novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and ...

  9. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    Directory of Open Access Journals (Sweden)

    V.I. Belozerov

    2016-12-01

    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  10. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  11. The CC-MGR. Combined Cycle - Modular Gas Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R.S. [Carlisle, Ontario (Canada)]. E-mail: rshart@cogeco.ca

    2007-07-01

    The Combined Cycle-Modular Gas Reactor (CC-MGR) takes advantage of established combined cycle gas turbine generation (CCGT) technology, utilizing a Modular High Temperature Gas Reactor (MHTGR) rather than natural gas to provide the heat source. Development of Helium cooled, graphite moderated High Temperature Gas Reactors (HTGRs) began with the Dragon project in the 1950s, and resulted in demonstration and commercial reactors being built in Germany (AVR-15 and THTR-300) and in the US (Peach Bottom 1 and Fort Saint Vrain). By the late 1980s all operating HTGRs were shut down and interest in the technology was fading. However, interest in HTGRs was revitalized over the past 15 years and HTGR research reactors now operating in Japan and in China and a commercial HTGR under construction in China. Two major MHTGR programmes currently in the design and development stage, the Pebble Bed Modular Reactor (PBMR) in South Africa, and the Gas Turbine-Modular Helium Reactor (GTMHR) by an international consortium headed by General Atomics, are focused on direct closed cycle technology in which the helium from the reactor is passed directly through a helium/gas turbine which subsequently drives a generator. In both of these designs, heat in the helium exhaust from the power turbine is transferred to the helium flow entering the reactor via a recuperator located downstream of the compressors. The amount of heat transferred in the recuperator is slightly greater than the reactor thermal power. In the CC-MGR power plant, the helium exhaust from the power turbine is directed to a steam generator which generates steam that subsequently drives a steam turbine-generator. The helium leaving the steam generator passes through a recuperator, where heat is transferred to the helium flow entering the reactor downstream of the compressors. This arrangement reduces the amount of heat transferred in the recuperator by approximately half, and results in a reduced reactor helium inlet temperature

  12. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report for period, 1 October 1977--31 December 1977

    Energy Technology Data Exchange (ETDEWEB)

    1978-03-20

    The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered includes the activities associated with the procurement of the materials for the screening test program and information from vendor certification for the materials received for the nuclear process heat candidate alloys. The design modifications to the helium purification system and the construction status of the simulated reactor helium supply system, testing equipment, and analysis instrumentation and equipment are discussed. Finally, the status and details of the data management are presented.

  13. Army Gas-Cooled Reactor Systems Program. ML-1 analytical design report. Volume II. Systems analysis: heat transfer and fluid flow

    Energy Technology Data Exchange (ETDEWEB)

    None

    1961-01-01

    The analysis preceding and supporting the design of the cooling system of the ML-1, a mobile, low-power, nuclear power plant, is described in sufficient detail for an engineer to follow the development of the design. Test results and similar data are used to support the calculations whenever possible.

  14. Liquid metal reactor air cooling baffle

    Science.gov (United States)

    Hunsbedt, A.

    1994-08-16

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat. 3 figs.

  15. Residence time distribution of the gas phase in a mechanically agitated gas-liquid reactor

    NARCIS (Netherlands)

    Thijert, M.P.G.; Oyevaar, M.H.; Kuper, W.J.; Westerterp, K.R.

    1992-01-01

    In this study we present a measuring method and extensive experimental data on the gas phase RTD in a mechanically agitated gas-liquid reactor with standard dimensions over a wide range of superficial gas velocities, agitation rates and agitator sizes. The results are modelled successfully, using

  16. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing of high-temperature gas-cooled reactor fuel containing U-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed.

  17. Experimental demonstration of relativistic electron cooling

    Energy Technology Data Exchange (ETDEWEB)

    Nagaitsev, S.; Broemmelsiek, D.; Burov, Alexey V.; Carlson, K.; Gattuso, C.; Hu, M.; Kazakevich, Grigory M.; Kroc, T.; Prost, L.; Pruss, S.; Sutherland, M.; Schmidt,; Seletskiy, S.; Shemyakin, A.; Tupikov, V.; Warner, A.; /Fermilab /Novosibirsk, IYF /Rochester U.

    2005-11-01

    We report on an experimental demonstration of electron cooling of high-energy antiprotons circulating in a storage ring. In our experiments, electron cooling, a well-established method at low energies (< 500 MeV/nucleon), was carried out in a new region of beam parameters, requiring a multi-MeV dc electron beam and an unusual beam transport line. In this letter we present the results of the longitudinal cooling force measurements and compare them with theoretical predictions.

  18. New concept of proliferation resistant sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Krivitski, I.Y.; Matveev, V.I.; Popov, E.P.; Savitski, V.I.; Tsikunov, A.G. [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-07-01

    The full text follows. It is proposed the concept of BN-800 sodium cooled fast reactor operating in the closed fuel cycle with special reprocessing technology. The use of nitride fuel allows improving the parameters of reactor safety (internal breeding {approx}1, zero value of sodium void reactivity effect), economy (one refueling per year), ecology (use of nitride enriched by nitrogen-15) and non-proliferation (use of reprocessing without separating the plutonium from uranium). The main difficulty of this type reactor development is that the technical project of BN-800 reactor with MOX fuel was developed. When using the nitride fuel it is necessary to serve (in max extent) the mail technical decisions of this project. This report presents first results on development and justification of the BN-800 reactor with nitride fuel core. (authors)

  19. Evaluation of gas cooling for pressurized phosphoric acid fuel cell stacks

    Science.gov (United States)

    Farooque, M.; Skok, A. J.; Maru, H. C.; Kothmann, R. E.; Harry, R. W.

    1983-01-01

    Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas cooling has already been demonstrated in atmospheric pressure stacks. This paper presents theoretical and experimental investigation of gas cooling for pressurized PAFC. Two approaches to gas cooling, Distributed Gas Cooling (DIGAS) and Separated Gas Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

  20. Ship exhaust gas plume cooling

    NARCIS (Netherlands)

    Schleijpen, H.M.A.; Neele, P.P.

    2004-01-01

    The exhaust gas plume is an important and sometimes dominating contributor to the infrared signature of ships. Suppression of the infrared ship signatures has been studied by TNO for the Royal Netherlands Navy over considerable time. This study deals with the suppression effects, which can be

  1. Liquid-cooled nuclear reactor. [Patent:; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Deinlein, H.; Kummer, G.

    1980-04-24

    H/sub 2/ is directly added to the coolant circuit. This requires a pipe bypassing the volume expansion tank and being connected with the suction side of the high pressure pump. The supply of H/sub 2/ is realized via ceramic filter catridges in a liquid cooled part of the pipe at the suction side of the high pressure pump. Thus, the danger of oxyhydrogen explosions is avoided.

  2. Development of Tritium Permeation Analysis Code and Tritium Transport in a High Temperature Gas-Cooled Reactor Coupled with Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2010-06-01

    Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.

  3. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor

    Science.gov (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  4. One-dimensional modeling of radial heat removal during depressurized heatup transients in modular pebble-bed and prismatic high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Savage, M.G.

    1984-07-01

    A one-dimensional computational model was developed to evaluate the heat removal capabilities of both prismatic-core and pebble-bed modular HTGRs during depressurized heatup transients. A correlation was incorporated to calculate the temperature- and neutron-fluence-dependent thermal conductivity of graphite. The modified Zehner-Schluender model was used to determine the effective thermal conductivity of a pebble bed, accounting for both conduction and radiation. Studies were performed for prismatic-core and pebble-bed modular HTGRs, and the results were compared to analyses performed by GA and GR, respectively. For the particular modular reactor design studied, the prismatic HTGR peak temperature was 2152.2/sup 0/C at 38 hours following the transient initiation, and the pebble-bed peak temperature was 1647.8/sup 0/C at 26 hours. These results compared favorably with those of GA and GE, with only slight differences caused by neglecting axial heat transfer in a one-dimensional radial model. This study found that the magnitude of the initial power density had a greater effect on the temperature excursion than did the initial temperature.

  5. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development program. Progress report, October 1, 1981-December 31, 1981. [Alloy-MA-956; alloy-MA-754

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, O.F.

    1982-06-15

    Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is descibed; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/, and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in the controlled-purity helium in the intensive screening test program is discussed. The results of metallographic studies of screening alloys exposed in controlled-purity helium for 3000 hours at 750/sup 0/C and 5500 hours at 950/sup 0/C, 3000 hours at 1050/sup 0/C and 6000 hours at 1050/sup 0/C and for weldments exposed in controlled-purity helium for 6000 hours at 750/sup 0/C and 6000 hours at 1050/sup 0/C are presented and discussed.

  6. Cooling and Heating Functions of Photoionized Gas

    Energy Technology Data Exchange (ETDEWEB)

    Gnedin, Nickolay Y.; /Chicago U., EFI /Fermilab /Chicago U., Astron. Astrophys. Ctr. /Chicago U., KICP; Hollon, Nicholas; /Chicago U., EFI /Chicago U., Astron. Astrophys. Ctr. /Chicago U., KICP

    2012-01-01

    Cooling functions of cosmic gas are a crucial ingredient for any study of gas dynamics and thermodynamics in the interstellar and intergalactic medium. As such, they have been studied extensively in the past under the assumption of collisional ionization equilibrium. However, for a wide range of applications, the local radiation field introduces a non-negligible, often dominant, modification to the cooling and heating functions. In the most general case, these modifications cannot be described in simple terms, and would require a detailed calculation with a large set of chemical species using a radiative transfer code (the well-known code Cloudy, for example). We show, however, that for a sufficiently general variation in the spectral shape and intensity of the incident radiation field, the cooling and heating functions can be approximated as depending only on (1) the photodissociation rate of molecular hydrogen, (2) the hydrogen photo-ionization rate, and (3) the photo-ionization rate of OVIII;more complex and more accurate approximations also exist. Such dependence is easy to tabulate and implement in cosmological or galactic-scale simulations, thus economically accounting for an important but rarely-included factor in the evolution of cosmic gas. We also show a few examples where the radiation environment has a large effect, the most spectacular of which is a quasar that suppresses gas cooling in its host halo without any mechanical or non-radiative thermal feedback.

  7. COOLING AND HEATING FUNCTIONS OF PHOTOIONIZED GAS

    Energy Technology Data Exchange (ETDEWEB)

    Gnedin, Nickolay Y. [Particle Astrophysics Center, Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Hollon, Nicholas, E-mail: gnedin@fnal.gov [Department of Astronomy and Astrophysics, University of Chicago, Chicago, IL 60637 (United States)

    2012-10-15

    Cooling and heating functions of cosmic gas are crucial ingredients for any study of gas dynamics and thermodynamics in the interstellar and intergalactic media. As such, they have been studied extensively in the past under the assumption of collisional ionization equilibrium. However, for a wide range of applications, the local radiation field introduces a non-negligible, often dominant, modification to the cooling and heating functions. In the most general case, these modifications cannot be described in simple terms and would require a detailed calculation with a large set of chemical species using a radiative transfer code (the well-known code Cloudy, for example). We show, however, that for a sufficiently general variation in the spectral shape and intensity of the incident radiation field, the cooling and heating functions can be approximated as depending only on several photoionization rates, which can be thought of as representative samples of the overall radiation field. This dependence is easy to tabulate and implement in cosmological or galactic-scale simulations, thus economically accounting for an important but rarely included factor in the evolution of cosmic gas. We also show a few examples where the radiation environment has a large effect, the most spectacular of which is a quasar that suppresses gas cooling in its host halo without any mechanical or non-radiative thermal feedback.

  8. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  9. Numerical and experimental investigation of turbine blade film cooling

    Science.gov (United States)

    Berkache, Amar; Dizene, Rabah

    2017-12-01

    The blades in a gas turbine engine are exposed to extreme temperature levels that exceed the melting temperature of the material. Therefore, efficient cooling is a requirement for high performance of the gas turbine engine. The present study investigates film cooling by means of 3D numerical simulations using a commercial code: Fluent. Three numerical models, namely k-ɛ, RSM and SST turbulence models; are applied and then prediction results are compared to experimental measurements conducted by PIV technique. The experimental model realized in the ENSEMA laboratory uses a flat plate with several rows of staggered holes. The performance of the injected flow into the mainstream is analyzed. The comparison shows that the RANS closure models improve the over-predictions of center-line film cooling velocities that is caused by the limitations of the RANS method due to its isotropy eddy diffusivity.

  10. Advanced-gas-cooled-nuclear-reactor materials evaluation and development program. Volume 1. Final report, September 23, 1976-September 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, O.F.

    1983-05-15

    Included in this report is a discussion of the materials selected for the screening phase and more intensive screening phase test programs and the systems and components for which they are candidate materials. Thirty-one (31) commercially available alloy and alloy/coating materials and ten (10) experimental alloys were evaluated in the program. The experimental test facilities developed as part of this program are discussed and experimental testing procedures are summarized. The results of the initial screening test programs are presented. This includes creep testing results and metallographic analyses of candidate materials exposed to simulated HTGR helium and air under stress at temperatures of 750/sup 0/, 850/sup 0/, 950/sup 0/, or 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/, or 1922/sup 0/F) for exposure times to 10,000 hours. Metallographic analyses, weight change and carbon analyses results, and post exposure room temperature tensile and Charpy V-notch impact test results are presented for candidate materials exposed unstressed under the conditions stated above.

  11. Army gas-cooled reactor systems program: an evaluation of the state-of-the-art of high-speed seals

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-12-01

    A literature survey was conducted to determine and evaluate the various types of seals which might be acceptable for use in the ML-1 t-c set and, based on this evaluation, to recommend the next logical steps in the seal development program. The survey indicated that six seal concepts exist which might satisfy the ML-1 air cycle requirements. Theoretical evaluation revealed that four of the concepts were sufficiently attractive to warrant experimental evaluation; the fifth did not appear acceptable and insufficient data existed to permit an evaluation of the sixth concept. The six concepts and summary comments concerning each are shown.

  12. Carbon Contained Ammonium Diuranate Gel Particles Preparation in Mid-process of High-temperature Gas-cooled Reactor Fuel Fabrication

    Directory of Open Access Journals (Sweden)

    Kyung Chai Jeong

    2016-02-01

    Full Text Available This study investigates the dispersibility of carbon in carbon contained ammonium diuranate (C-ADU gel particles and the characteristics of C-ADU gel liquid droplets produced by the vibrating nozzle and integrated aging–washing–drying equipment. It was noted that the excellent stability of carbon dispersion was only observed in the C-ADU gel particle that contained carbon black named CB 10. ADU gel liquid droplets containing carbon particles with the excellent sphericity of approximately 1,950 μm were then obtained using an 80–100-Hz vibrating nozzle system. Dried C-ADU gel particles obtained by the aging–washing–drying equipment were thermal decomposed until 500°C at a rate of 1°C/min in an air or in 4% H2 gas atmosphere. The thermally decomposed C-ADU gel particles showed 24% weight loss and a more complicated profile than that of ADU gel particles.

  13. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  14. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  15. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  16. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  17. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  18. Cooled membrane for high sensitivity gas sampling.

    Science.gov (United States)

    Jiang, Ruifen; Pawliszyn, Janusz

    2014-04-18

    A novel sample preparation method that combines the advantages of high surface area geometry and cold surface effect was proposed to achieve high sensitivity gas sampling. To accomplish this goal, a device that enables the membrane to be cooled down was developed for sampling, and a gas chromatograph-mass spectrometer was used for separation and quantification analysis. Method development included investigation of the effect of membrane temperature, membrane size, gas flow rate and humidity. Results showed that high sensitivity for equilibrium sampling, such as limonene sampling in the current study could be achieved by either cooling down the membrane and/or using a large volume extraction phase. On the other hand, for pre-equilibrium extraction, in which the extracted amount was mainly determined by membrane surface area and diffusion coefficient, high sensitivity could be obtained by using thinner membranes with a larger surface and/or a higher sampling flow rate. In addition, humidity showed no significant influence on extraction efficiency, due to the absorption property of the liquid extraction phase. Next, the limit of detection (LOD) was found, and the reproducibility of the developed cooled membrane gas sampling method was evaluated. Results showed that LODs with a membrane diameter of 19mm at room temperature sampling were 9.2ng/L, 0.12ng/L, 0.10ng/L for limonene, cinnamaldehyde and 2-pentadecanone, respectively. Intra- and inter-membrane sampling reproducibility revealed RSD% lower than 8% and 13%, respectively. Results uniformly demonstrated that the proposed cooled membrane device could serve as an alternative powerful tool for future gas sampling. Copyright © 2014 Elsevier B.V. All rights reserved.

  19. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  20. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  1. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  2. Axial Dispersion and Back-mixing of Gas Phase in Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rahman Al-Musafir

    2013-04-01

    Full Text Available Despite the worldwide attended of pebble bed reactors (PBRs, there is a lack of fundamental understanding of the complex flow pattern. In this work, the non-ideal flow behavior of the gas phase which is used for cooling has been investigated experimentally in a 0.3 m diameter pebble bed. The extent of mixing and dispersion of the gas phase has been qualified. The effect of gas velocity on the axial dispersion has been investigated with range from 0.05 to 0.6 m/s covering both the laminar and turbulent flow regimes. Glass bead particles of 1.2 cm diameter and 2.5 gm/cm3 which is randomly and closely packed have been used to mimic the pebbles. An advanced gas tracer technique was applied to measure the residence time distribution (RTD of gas phase using impulse tracer. The axial dispersion coefficients of gas phase in the studied pebble bed have been estimated using the axial dispersion model (ADM. It was found that the flow pattern of the gas phase deviates from plug flow depending on the superficial gas velocity. The results showed that the dispersion of the gas reduces as the gas velocity and Reynolds numbers increased.

  3. Neutronics of a mixed-flow gas-core reactor

    Energy Technology Data Exchange (ETDEWEB)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF/sub 6/ (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation.

  4. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  5. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  6. Aerosol flow in a tube furnace reactor of gas-phase synthesised silver nanoparticles

    Science.gov (United States)

    Mitrakos, D.; Jokiniemi, J.; Backman, U.; Housiadas, C.

    2008-12-01

    In a previous work, gas-phase synthesis of silver nanoparticles through evaporation of silver powder and subsequent particle nucleation by cooling was shown to be a viable method for achieving high purity silver nanoparticles (Backman et al. J Nanopart Res 4:325-335, 2002). In order to control the size of the produced nanoparticles, careful design of the reactor is required with respect to thermal and flow characteristics. In the present work, the silver nanoparticle reactor is rigorously simulated by means of multidimensional computational fluid and particle dynamics. The CFD-computed flow is input for a combined simulation of the vapour field and particle homogeneous nucleation, growth and coagulation. The results are compared with the experimental data and with the predictions from the usually employed simple model of an idealized plug flow reactor. The multidimensional CFD-based analysis is shown to explain and help understand different aspects of the reactor operation and size distribution of the particles produced. Yet the simple plug flow method is found to provide reasonable accuracy when an appropriate correction factor is used for the nucleation rate. Considering its robustness and computational simplicity, the plug flow method can be qualified as adequate from the engineering practical point of view for the case of silver nanoparticle reactors.

  7. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned.

  8. Numerical approach for quantification of selfwastage phenomena in sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Sunghyon Jang

    2015-10-01

    Full Text Available Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium–water reaction (SWR takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called “self-wastage phenomena.” A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium–water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM. The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm.

  9. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  10. Experimental Investigation of Effect on Hydrate Formation in Spray Reactor

    Directory of Open Access Journals (Sweden)

    Jianzhong Zhao

    2015-01-01

    Full Text Available The effects of reaction condition on hydrate formation were conducted in spray reactor. The temperature, pressure, and gas volume of reaction on hydrate formation were measured in pure water and SDS solutions at different temperature and pressure with a high-pressure experimental rig for hydrate formation. The experimental data and result reveal that additives could improve the hydrate formation rate and gas storage capacity. Temperature and pressure can restrict the hydrate formation. Lower temperature and higher pressure can promote hydrate formation, but they can increase production cost. So these factors should be considered synthetically. The investigation will promote the advance of gas storage technology in hydrates.

  11. Cooling an Optically Trapped Ultracold Fermi Gas by Periodical Driving.

    Science.gov (United States)

    Li, Jiaming; de Melo, Leonardo F; Luo, Le

    2017-03-30

    We present a cooling method for a cold Fermi gas by parametrically driving atomic motions in a crossed-beam optical dipole trap (ODT). Our method employs the anharmonicity of the ODT, in which the hotter atoms at the edge of the trap feel the anharmonic components of the trapping potential, while the colder atoms in the center of the trap feel the harmonic one. By modulating the trap depth with frequencies that are resonant with the anharmonic components, we selectively excite the hotter atoms out of the trap while keeping the colder atoms in the trap, generating parametric cooling. This experimental protocol starts with a magneto-optical trap (MOT) that is loaded by a Zeeman slower. The precooled atoms in the MOT are then transferred to an ODT, and a bias magnetic field is applied to create an interacting Fermi gas. We then lower the trapping potential to prepare a cold Fermi gas near the degenerate temperature. After that, we sweep the magnetic field to the noninteracting regime of the Fermi gas, in which the parametric cooling can be manifested by modulating the intensity of the optical trapping beams. We find that the parametric cooling effect strongly depends on the modulation frequencies and amplitudes. With the optimized frequency and amplitude, we measure the dependence of the cloud energy on the modulation time. We observe that the cloud energy is changed in an anisotropic way, where the energy of the axial direction is significantly reduced by parametric driving. The cooling effect is limited to the axial direction because the dominant anharmonicity of the crossed-beam ODT is along the axial direction. Finally, we propose to extend this protocol for the trapping potentials of large anharmonicity in all directions, which provides a promising scheme for cooling quantum gases using external driving.

  12. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  13. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  14. Feasibility analysis of two-phase MHD energy conversion for liquid metal cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wu Qiao [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, Corvallis, OR 97331 (United States)], E-mail: qiao@engr.orst.edu; Schubring, DuWayne L. [Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, Corvallis, OR 97331 (United States); Sienicki, James J. [Reactor Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States)

    2007-11-15

    A two-phase MHD energy conversion unit is proposed to a liquid metal cooled fast reactor. Using supercritical CO{sub 2} as the working fluid in the gas cycle without considering friction and heat losses, the optimized cycles efficiency is obtained, which is about 5% higher than that of the gas turbine Brayton cycle with the same regenerator/compressor configurations. Based on a simple MHD power analysis and the two-phase homogeneous flow model, the important system operational conditions were estimated. The results suggest that a liquid lead pump of at least 20% of the MHD power output is needed in order to convert the 400 MW reactor heat into electricity at the specified thermal efficiency, unless a mixture foam flow of void fraction greater than 80% is achievable at very high mixture velocity.

  15. Investigation of mixing chamber for experimental FGD reactor

    Directory of Open Access Journals (Sweden)

    Novosád Jan

    2017-01-01

    Full Text Available This article deals with numerical investigation of flow and mixing of air and sulphur dioxide SO2 in designated mixing chamber. The mixing chamber is a part of experimental laboratory reactor designed for simulating the flue gas desulfurization (FGD process. Aim of this work is the numerical investigation of effect of different mixing chamber geometries to mixture composition, especially to mass fraction of sulphur dioxide. Using of similar concentration of sulphur dioxide in the experimental reactor as in the real process is necessary to be able to make additional research. Conclusion describes the effect of different geometries of mixing chamber to mixing. The aim of this work is to develop perfectly works mixing chamber, which will be manufactured and then implemented into experimental FGD reactor. The results will be validated by experiment after the mixing chamber will be manufactured.

  16. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  17. Lead-cooled fast reactor (LFR) overview and perspectives

    OpenAIRE

    CINOTTI Luciano; Smith, Craig F.; SEKIMOTO, HIROSHI

    2009-01-01

    The GIF Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology with great potential to meet the small-unit electricity needs of remote sites while also offering advantages as a large system for grid-connected power stations. The LFR features a fast- neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner of minor actinides from spent fuel and as a burner/breeder. An important feature of the LFR is the ...

  18. Stagnation Point Heat Transfer with Gas Injection Cooling

    Science.gov (United States)

    Vancrayenest, B.; Tran, M. D.; Fletcher, D. G.

    2005-01-01

    The present paper deals with an experimental study of the stagnation-point heat transfer to a cooled copper surface with gas injection under subsonic conditions. Test were made with a probe that combined a steady-state water-cooled calorimeter that allows the capability to study convective blockage and to perform heat transfer measurements in presence of gas injection in the stagnation region. The copper probe was pierced by 52 holes, representing 2.4% of the total probe surface. The 1.2 MW high enthalpy plasma wind tunnel was operated at anode powers between 130 and 230 kW and a static pressures from 35 hPa up to 200 hPa. Air, carbon dioxide and argon were injected in the mass flow range 0-0.4 g/s in the boundary layer developed around the 50 mm diameter probe. The measured stagnation-point heat transfer rates are reported and discussed.

  19. Passive modular gas safety system for a reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Isaev, I.F.; Kulakov, A.A.; Sivokon, V.P.; Udovenko, A.N.; Ionaitis, R.R.

    1994-01-01

    Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

  20. A reverse flow catalytic membrane reactor for the production of syngas: an experimental study

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    In this paper experimental results are presented for a demonstration unit of a recently proposed novel integrated reactor concept (Smit et. al., 2005) for the partial oxidation of natural gas to syngas (POM), namely a Reverse Flow Catalytic Membrane Reactor (RFCMR). Natural gas has great potential

  1. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  2. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  3. Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

    Energy Technology Data Exchange (ETDEWEB)

    Murav’ev, V. P., E-mail: murval1@mail.ru

    2016-07-15

    The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

  4. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    Directory of Open Access Journals (Sweden)

    P.M. Udiyani

    2017-02-01

    Full Text Available Experimental power reactor (RDE which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated using ORIGEN-2 whose library has been modified for HTGR type, and the assumptions of defects of the TRISO fuel and release fraction for each compartment of RDE safety system used a reference parameter. The results showed that the important source terms of RDE are group of noble gases (Kr and Xe, halogen (I, Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTGR source terms with the same power. Keywords: routine discharge, radionuclide, source term, RDE, HTGR

  5. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    OpenAIRE

    Zhao, Pengcheng; Shi, Kangli; Li, Shuzhou; Feng, Jingchao; Chen, Hongli

    2016-01-01

    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  6. Gas tagging and cover gas combination for nuclear reactor

    Science.gov (United States)

    Gross, Kenny C.; Laug, Matthew T.

    1985-01-01

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  7. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

  8. Experimental Progress in Fast Cooling in the ESR

    CERN Document Server

    Steck, Markus; Beller, Peter; Franzke, Bernhard; Nolden, Fritz

    2005-01-01

    The ESR storage ring at GSI is operated with highly charged heavy ions. Due to the high electric charge the ions interact much stronger with electromagnetic fields. Therefore both cooling methods which are applied to stored ions in the ESR, stochastic cooling and electron cooling, are more powerful than for singly charged particles. The experimental results exhibit cooling times for stochastic cooling of a few seconds. For cold ion beams, electron cooling provides cooling times which are one to two orders of magnitude smaller. The beams are cooled to beam parameters which are limited by intrabeam scattering. At small ion numbers, however, intrabeam scattering is suppressed by electron cooling, clear evidence was found that the ion beam forms a one-dimensional ordered structure, a linear chain of ions. The strengths of stochastic cooling and electron cooling are complementary and can be combined favorably. Stochastic cooling is employed for pre-cooling of hot secondary beams followed by electron cooling to pro...

  9. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  10. Compressed Gas Safety for Experimental Fusion Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2004-09-01

    Experimental fusion facilities present a variety of hazards to the operators and staff. There are unique or specialized hazards, including magnetic fields, cryogens, radio frequency emissions, and vacuum reservoirs. There are also more general industrial hazards, such as a wide variety of electrical power, pressurized air, and cooling water systems in use, there are crane and hoist loads, working at height, and handling compressed gas cylinders. This paper outlines the projectile hazard assoicated with compressed gas cylinders and mthods of treatment to provide for compressed gas safety. This information should be of interest to personnel at both magnetic and inertial fusion experiments.

  11. Evaluation of the Gas Turbine Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  12. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor; Amelioration des caracteristiques de la dissipation de la chaleur de decroissance pour les reacteurs a neutrons rapides de quatrieme generation refroidi au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.S.

    2010-09-07

    The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For de-pressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research - aimed at achieving enhanced passivity of the DHR system for the GFR - are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study - TRACE and CATHARE - are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a

  13. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  14. Cooled gas turbine blade edge flow analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mendonca, Marcio Teixeira de [Instituto Tecnologico de Aeronautica, Divisao de Engenharia Mecanica Aeronautica ITA/IEM, Sao Jose dos Campos, SP (Brazil)], e-mail: marcio@ita.br

    2010-07-01

    The flow on the rotating blades of a turbine is unsteady due to the wake of the stator blade row upstream. This unsteadiness is a source of losses and complex flow structures on the rotor blade due to the variation on the turbulence levels and location of the boundary layer laminar to turbulent transition. Convective cooled blades often time have cooling air ejected at the trailing edge right at the blade wake. The present investigation presents an analysis of a canonical flow consistent with the flow topology found at the trailing edge of a gas turbine blade with coolant ejection. A hydrodynamic stability analysis is performed for the combined wake and jet velocity profiles given by a gaussian distribution representing the turbulent rms wake and a laminar jet superposed. The growth rate of any instability found on the flow is an indication of faster mixing, resulting in a reduction on the wake velocity defect and consequently on the complexity associated with it. The results show that increasing the Mach number or the three-dimensionality of the disturbances result in a reduction of the amplification rate. When the flow at the trailing edge is modified by a jet, the amplification rates are lower, but the range of unstable stream wise wavenumbers is larger. (author)

  15. Steam cooling system for a gas turbine

    Science.gov (United States)

    Wilson, Ian David; Barb, Kevin Joseph; Li, Ming Cheng; Hyde, Susan Marie; Mashey, Thomas Charles; Wesorick, Ronald Richard; Glynn, Christopher Charles; Hemsworth, Martin C.

    2002-01-01

    The steam cooling circuit for a gas turbine includes a bore tube assembly supplying steam to circumferentially spaced radial tubes coupled to supply elbows for transitioning the radial steam flow in an axial direction along steam supply tubes adjacent the rim of the rotor. The supply tubes supply steam to circumferentially spaced manifold segments located on the aft side of the 1-2 spacer for supplying steam to the buckets of the first and second stages. Spent return steam from these buckets flows to a plurality of circumferentially spaced return manifold segments disposed on the forward face of the 1-2 spacer. Crossover tubes couple the steam supply from the steam supply manifold segments through the 1-2 spacer to the buckets of the first stage. Crossover tubes through the 1-2 spacer also return steam from the buckets of the second stage to the return manifold segments. Axially extending return tubes convey spent cooling steam from the return manifold segments to radial tubes via return elbows.

  16. Metabolic modeling of synthesis gas fermentation in bubble column reactors.

    Science.gov (United States)

    Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

    2015-01-01

    A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

  17. Gas turbine vane cooling air insert

    Energy Technology Data Exchange (ETDEWEB)

    North, W.E.; Hultgren, K.G.; Dishman, C.D.; Van Heusden, G.S.

    1992-09-08

    This patent describes a gas turbine. It comprises turbine vanes, each of the vanes supplied with cooling air and having: an airfoil portion forming a first cavity having an insert disposed therein for directing the flow of the cooling air, the insert having first and second insert ends; a shroud portion from which the airfoil portion extends, the insert attached to the shroud portion at the first insert end; an insert extension extending through a portion of the insert and extending beyond the first insert end, the insert extension and the insert forming an annular gap therebetween separating the insert from the insert extension; a plate covering at least a portion of the shroud, the plate having a first hole formed therein through which the insert extension extends; and at least a first seal extending between the insert extension and the insert, and sealing the annular gap therebetween. This patent also describes a method of making a gas turbine. It comprises welding a first tubular insert adjacent its first end to a vane outer shroud; partially inserting a second tubular insert into the first tubular member and attaching the second tubular insert thereto; placing a plate having a hole formed therein on the outer shroud so that the hole surrounds the second tubular insert; and attaching the second tubular insert to the plate by placing a first seal between the first and second tubular inserts and attaching the first seal to each of the first and second tubular inserts, and placing a second seal between the second tubular insert and the plate and welding the second seal to the second tubular insert and the plate.

  18. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  19. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  20. Experimental evaluation of cooling efficiency of the high performance cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Nemec, Patrik, E-mail: patrik.nemec@fstroj.uniza.sk; Malcho, Milan, E-mail: milan.malcho@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engineering, Univerzitna 1, 010 26 Žilina (Slovakia)

    2016-06-30

    This work deal with experimental evaluation of cooling efficiency of cooling device capable transfer high heat fluxes from electric elements to the surrounding. The work contain description of cooling device, working principle of cooling device, construction of cooling device. Experimental part describe the measuring method of device cooling efficiency evaluation. The work results are presented in graphic visualization of temperature dependence of the contact area surface between cooling device evaporator and electronic components on the loaded heat of electronic components in range from 250 to 740 W and temperature dependence of the loop thermosiphon condenser surface on the loaded heat of electronic components in range from 250 to 740 W.

  1. Experimental and numerical simulation of passive decay heat removal by sump cooling after cool melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U.; Kuhn, D.; Mueller, U. [Institut fuer Angewandet Thermo- und Fluiddynamik (IATF) (Germany)

    1997-12-01

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase and two-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software package Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a first statement with regard to the feasibility of the sump cooling concept. 11 refs., 9 figs., 3 tabs.

  2. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of fuel and claddings during accident are still below limitations which are in secure condition.

  3. Improved gas tagging and cover gas combination for nuclear reactor

    Science.gov (United States)

    Gross, K.C.; Laug, M.T.

    1983-09-26

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  4. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Osinga, T. [ETH-Zuerich (Switzerland); Olalde, G. [CNRS Odeillo (France); Steinfeld, A. [PSI and ETHZ (Switzerland)

    2005-03-01

    A numerical model is formulated for the SOLZINC solar chemical reactor for the production of Zn by carbothermal reduction of ZnO. The model involves solving, by the finite-volume technique, a 1D unsteady state energy equation that couples heat transfer to the chemical kinetics for a shrinking packed bed exposed to thermal radiation. Validation is accomplished by comparison with experimentally measured temperature profiles and Zn production rates as a function of time, obtained for a 5-kW solar reactor tested at PSI's solar furnace. (author)

  5. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  6. Gas turbine heat transfer and cooling technology

    CERN Document Server

    Han, Je-Chin; Ekkad, Srinath

    2012-01-01

    FundamentalsNeed for Turbine Blade CoolingTurbine-Cooling TechnologyTurbine Heat Transfer and Cooling IssuesStructure of the BookReview Articles and Book Chapters on Turbine Cooling and Heat TransferNew Information from 2000 to 2010ReferencesTurbine Heat TransferIntroductionTurbine-Stage Heat TransferCascade Vane Heat-Transfer ExperimentsCascade Blade Heat TransferAirfoil Endwall Heat TransferTurbine Rotor Blade Tip Heat TransferLeading-Edge Region Heat TransferFlat-Surface Heat TransferNew Information from 2000 to 20102.10 ClosureReferencesTurbine Film CoolingIntroductionFilm Cooling on Rotat

  7. Entropy generation in a channel resembling gas turbine cooling ...

    Indian Academy of Sciences (India)

    Abstract. Flow into a passage resembling a gas turbine blade cooling passage is considered and entropy .... for the flow systems associated with the cooling applications. In the present study, rectangular .... Since we are using ideal gas law to incorporate the density variation with temperature, the code does not permit use of ...

  8. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  9. Evaluating the income and employment impacts of gas cooling technologies

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J. [Oak Ridge National Lab., TN (United States); Laitner, S.

    1995-03-01

    The purpose of this study is to estimate the potential employment and income benefits of the emerging market for gas cooling products. The emphasis here is on exports because that is the major opportunity for the U.S. heating, ventilating, and air-conditioning (HVAC) industry. But domestic markets are also important and considered here because without a significant domestic market, it is unlikely that the plant investments, jobs, and income associated with gas cooling exports would be retained within the United States. The prospects for significant gas cooling exports appear promising for a variety of reasons. There is an expanding need for cooling in the developing world, natural gas is widely available, electric infrastructures are over-stressed in many areas, and the cost of building new gas infrastructure is modest compared to the cost of new electric infrastructure. Global gas cooling competition is currently limited, with Japanese and U.S. companies, and their foreign business partners, the only product sources. U.S. manufacturers of HVAC products are well positioned to compete globally, and are already one of the faster growing goods-exporting sectors of the U.S. economy. Net HVAC exports grew by over 800 percent from 1987 to 1992 and currently exceed $2.6 billion annually (ARI 1994). Net gas cooling job and income creation are estimated using an economic input-output model to compare a reference case to a gas cooling scenario. The reference case reflects current policies, practices, and trends with respect to conventional electric cooling technologies. The gas cooling scenario examines the impact of accelerated use of natural gas cooling technologies here and abroad.

  10. Gas Mixtures for Welding with Micro-Jet Cooling

    Directory of Open Access Journals (Sweden)

    Węgrzyn T.

    2015-04-01

    Full Text Available Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechanical properties of weld was presented in terms of various gas mixtures selection for micro-jet cooling.

  11. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Science.gov (United States)

    2010-01-01

    ...-Cooled Power Reactors J Appendix J to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor Containment... basis accident and specified either in the technical specification or associated bases. J. Pt (p.s.i.g...

  12. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized.

  13. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  14. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    Science.gov (United States)

    Galvez, Cristhian

    2011-12-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  15. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.'' DG-1277...

  16. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  17. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  18. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  19. Novel Applications of Buffer-gas Cooling to Cold Atoms, Diatomic Molecules, and Large Molecules

    Science.gov (United States)

    Drayna, Garrett Korda

    Cold gases of atoms and molecules provide a system for the exploration of a diverse set of physical phenomena. For example, cold gasses of magnetically and electrically polar atoms and molecules are ideal systems for quantum simulation and quantum computation experiments, and cold gasses of large polar molecules allow for novel spectroscopic techniques. Buffer-gas cooling is a robust and widely applicable method for cooling atoms and molecules to temperatures of approximately 1 Kelvin. In this thesis, I present novel applications of buffer-gas cooling to obtaining gases of trapped, ultracold atoms and diatomic molecules, as well as the study of the cooling of large organic molecules. In the first experiment of this thesis, a buffer-gas beam source of atoms is used to directly load a magneto-optical trap. Due to the versatility of the buffer-gas beam source, we obtain trapped, sub-milliKelvin gases of four different lanthanide species using the same experimental apparatus. In the second experiment of this thesis, a buffer-gas beam is used as the initial stage of an experiment to directly laser cool and magneto-optically trap the diatomic molecule CaF. In the third experiment of this thesis, buffer-gas cooling is used to study the cooling of the conformational state of large organic molecules. We directly observe conformational relaxation of gas-phase 1,2-propanediol due to cold collisions with helium gas. Lastly, I present preliminary results on a variety of novel applications of buffer-gas cooling, such as mixture analysis, separation of chiral mixtures, the measurement of parity-violation in chiral molecules, and the cooling and spectroscopy of highly unstable reaction intermediates.

  20. Studying the properties and behaviour of high-level radioactive wastes from the BOR-60 reactor U-Pu and U spent fuel experimental gas-fluoride reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Kirillovich, A.P.; Lavrinovich, Yu.G.; Vorobej, M.P.; Pimonov, Yu.I.

    1982-07-01

    The results of investigations of physical-chemical and radiation properties of fluoride radioactive wastes produced during experimental reprocessing of spent oxide uranium-plutonium fuel as well as high-level radioactive waste behaviour in the process of their six-year controlled storage are presented. Radioactive gas release from solid wastes, gaseous phase composition, radionuclide leaching rate are determined. Investigations are performed at a special bench. Energy release of the spent fuel and high-level radioactive wastes is determined by means of heat-conducting type calorimeter. Gas mixture composition in containers with wastes is determined by mass-spectrometric method at equilibrium temperature of high-level radioactive product self-heating and at external container heating up to 700 deg C. Thermal-physical characteristics of solid fluoride wastes are found by the differential thermography method under quasistationary heating. The results obtained show that about a half (44.8-60.9%) of fission product radioactivity is concentrated in fluorination wastes, specific heat release of which constitutes 50-52 W/kg, while ..beta..-activity exceeds 550 TBq/kg. Main contribution into ..beta..-activity is made by Ce, /sup 144/Pr, Ru, /sup 106/Rh, Zr, /sup 95/Nb, /sup 137/Cs. With waste storage time increase their thermal stability increases. It is concluded that the investigation results can be used for calculating the conditions of safe storage of high-level radiactive solid fluoride wastes and optimization of the technological process of spent fuel gas-fluoride reprossing.

  1. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov; Chichester, H.J.M.; Medvedev, P.G.; Hayes, S.L.; Teague, M.C.

    2015-10-15

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  2. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K. J.; Park, C. K.; Seok, S. D.; Park, R. J.; Yi, S. J.; Kang, K. H.; Ham, Y. S.; Cho, Y. R.; Kim, J. H.; Jeong, J. H.; Shin, K. Y.; Cho, J. S.; Kim, D. H.

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  3. Improvement of Cooling Technology through Atmosphere Gas Management

    Energy Technology Data Exchange (ETDEWEB)

    Renard, Michel; Dosogne, Edgaar; Crutzen, Jean Pierre; Raick, Jean Mare [DREVER INTERNATIONAL S.A., Liege (Belgium); Ji, Ma Jia; Jun, Lv; Zhi, Ma Bing [SHOUGANG Cold Rolling Mill Headquarter, Beijin (China)

    2009-12-15

    The production of advanced high strength steels requires the improvement of cooling technology. The use of high cooling rates allows relatively low levels of expensive alloying additions to ensure sufficient hardenability. In classical annealing and hot-dip galvanizing lines a mixing station is used to provide atmosphere gas containing 3-5% hydrogen and 97-95% nitrogen in the various sections of the furnace, including the rapid cooling section. Heat exchange enhancement in this cooling section can be insured by the increased hydrogen concentration. Driver international developed a patented improvement of cooling technology based on the following features: pure hydrogen gas is injected only in the rapid cooling section whereas the different sections of the furnace are supplied with pure nitrogen gas: the control of flows through atmosphere gas management allows to get high hydrogen concentration in cooling section and low hydrogen content in the other furnace zones. This cooling technology development insures higher cooling rates without additional expensive hydrogen gas consumption and without the use of complex sealing equipment between zones. In addition reduction in electrical energy consumption is obtained. This atmosphere control development can be combined with geometrical design improvements in order to get optimised cooling technology providing high cooling rates as well as reduced strip vibration amplitudes. Extensive validation of theoretical research has been conducted on industrial lines. New lines as well as existing lines, with limited modifications, can be equipped with this new development. Up to now this technology has successfully been implemented on 6 existing and 7 new lines in Europe and Asia.

  4. MASS-TRANSFER IN GAS-LIQUID SLURRY REACTORS

    NARCIS (Netherlands)

    BEENACKERS, AACM; VANSWAAIJ, WPM

    A critical review is presented on the mass transfer characteristics of gas-liquid slurry reactors. The recent findings on the influence of the presence of solid particles on the following mass transfer parameters in slurry reactors are discussed: volumetric gas-liquid mass transfer coefficients

  5. Mass transfer in gas-liquid slurry reactors

    NARCIS (Netherlands)

    Beenackers, A.A.C.M.; van Swaaij, Willibrordus Petrus Maria

    1993-01-01

    A critical review is presented on the mass transfer characteristics of gas¿liquid slurry reactors. The recent findings on the influence of the presence of solid particles on the following mass transfer parameters in slurry reactors are discussed: volumetric gas¿liquid mass transfer coefficients

  6. Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

    Science.gov (United States)

    Lantz, E.; Breitwieser, R.; Niederauer, G. F.

    1974-01-01

    There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

  7. Optimization of Internal Cooling Fins for Metal Hydride Reactors

    Directory of Open Access Journals (Sweden)

    Vamsi Krishna Kukkapalli

    2016-06-01

    Full Text Available Metal hydride alloys are considered as a promising alternative to conventional hydrogen storage cylinders and mechanical hydrogen compressors. Compared to storing in a classic gas tank, metal hydride alloys can store hydrogen at nearly room pressure and use less volume to store the same amount of hydrogen. However, this hydrogen storage method necessitates an effective way to reject the heat released from the exothermic hydriding reaction. In this paper, a finned conductive insert is adopted to improve the heat transfer in the cylindrical reactor. The fins collect the heat that is volumetrically generated in LaNi5 metal hydride alloys and deliver it to the channel located in the center, through which a refrigerant flows. A multiple-physics modeling is performed to analyze the transient heat and mass transfer during the hydrogen absorption process. Fin design is made to identify the optimum shape of the finned insert for the best heat rejection. For the shape optimization, use of a predefined transient heat generation function is proposed. Simulations show that there exists an optimal length for the fin geometry.

  8. Improving activity transport models for water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Burrill, K.A

    2001-08-01

    Eight current models for describing radioactivity transport and radiation field growth around water-cooled nuclear power reactors have been reviewed and assessed. A frequent failing of the models is the arbitrary nature of the determination of the important processes. Nearly all modelers agree that the kinetics of deposition and release of both dissolved and particulate material must be described. Plant data must be used to guide the selection and development of suitable improved models, with a minimum of empirically-based rate constraints being used. Limiting case modelling based on experimental data is suggested as a way to simplify current models and remove their subjectivity. Improved models must consider the recent change to 'coordinated water chemistry' that appears to produce normal solubility behaviour for dissolved iron throughout the fuel cycle in PWRs, but retrograde solubility remains for dissolved nickel. Profiles are suggested for dissolved iron and nickel concentrations around the heat transport system in CANDU reactors, which operate nominally at constant chemistry, i.e., pH{sub T} constant with time, and which use carbon steel isothermal piping. These diagrams are modified for a CANDU reactor with stainless steel piping, in order to show the changes expected. The significance of these profiles for transport in PWRs is discussed for further model improvement. (author)

  9. Heat Transfer and Cooling in Gas Turbines

    Science.gov (United States)

    1985-09-01

    the detailed component internal heat transfer for a variety of families of cooling schemes, and (c) to choose from among and withir those families to...1965. 32. Metzger, D.E., and Grochowsky, 1.D., "Heat Transfer Between an Impinging Jet and a Rotating Dink ," J. Heat Tranafer, Trans. ASME, 99, pp. 663

  10. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Gas Mixtures for Welding with Micro-Jet Cooling

    OpenAIRE

    Węgrzyn T.

    2015-01-01

    Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechani...

  12. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors...

  13. Experimental and statistical investigation of thermally induced failure in reactor fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Lunsford, J.L.; Imprescia, R.J.; Bowman, A.L.; Radosevich, C.E.

    1980-10-01

    An incomplete experimental study into the failure statistics of fuel particle for the high-temperature gas-cooled reactor (HTGR) is described. Fuel particles failure was induced by thermal ramping from room temperature to temperatures in the vicinity of 2273/sup 0/K to 2773/sup 0/K in 2 to 30 h and detected by the appearance of /sup 85/Kr in the helium carrier gas used to sweep the furnace. The concentration of krypton, a beta emitter, was detected by measuring the current that resulted when the helium sweep gas was passed through an ionization chamber. TRISO fuel particles gave a krypton concentration profile as a function of time that built up in several minutes and decayed in a fraction of an hour. This profile, which was temperature independent, was similar to the impulse response of the ionization chamber, suggesting that the TRISO particles failed instantaneously and completely. BISO fuel particles gave a krypton concentration profile as a function of time that built up in a fraction of an hour and decayed in a fraction of a day. This profile was strongly temperature dependent, suggesting that krypton release was diffusion controlled, i.e., that the krypton was diffusing through a sound coat, or that the BISO coating failed but that the krypton was unable to escape the kernel without diffusion, or that a combination of pre- and postfailure diffusion accompanied partial or complete failure.

  14. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables.

  15. Influence of collector heat capacity and internal conditions of heat exchanger on cool-down process of small gas liquefier

    Science.gov (United States)

    Saberimoghaddam, Ali; Bahri Rasht Abadi, Mohammad Mahdi

    2018-01-01

    Joule-Thomson cooling systems are commonly used in gas liquefaction. In small gas liquefiers, transient cool-down time is high. Selecting suitable conditions for cooling down process leads to decrease in time and cost. In the present work, transient thermal behavior of Joule-Thomson cooling system including counter current helically coiled tube in tube heat exchanger, expansion valve, and collector was studied using experimental tests and simulations. The experiments were performed using small gas liquefier and nitrogen gas as working fluid. The heat exchanger was thermally studied by experimental data obtained from a small gas liquefier. In addition, the simulations were performed using experimental data as variable boundary conditions. A comparison was done between presented and conventional methods. The effect of collector heat capacity and convection heat transfer coefficient inside the tubes on system performance was studied using temperature profiles along the heat exchanger.

  16. Compatibility of gas turbine materials with steam cooling

    Energy Technology Data Exchange (ETDEWEB)

    Desai, V.; Tamboli, D.; Patel, Y. [Univ. of Central Florida, Orlando, FL (United States)

    1995-10-01

    Gas turbines had been traditionally used for peak load plants and remote locations as they offer advantage of low installation costs and quick start up time. Their use as a base load generator had not been feasible owing to their poor efficiency. However, with the advent of gas turbines based combined cycle plants (CCPs), continued advances in efficiency are being made. Coupled with ultra low NO{sub x} emissions, coal compatibility and higher unit output, gas turbines are now competing with conventional power plants for base load power generation. Currently, the turbines are designed with TIT of 2300{degrees}F and metal temperatures are maintained around 1700{degrees}F by using air cooling. New higher efficiency ATS turbines will have TIT as high as 2700{degrees}F. To withstand this high temperature improved materials, coatings, and advances in cooling system and design are warranted. Development of advanced materials with better capabilities specifically for land base applications are time consuming and may not be available by ATS time frame or may prove costly for the first generation ATS gas turbines. Therefore improvement in the cooling system of hot components, which can take place in a relatively shorter time frame, is important. One way to improve cooling efficiency is to use better cooling agent. Steam as an alternate cooling agent offers attractive advantages because of its higher specific heat (almost twice that of air) and lower viscosity.

  17. Design of conduction cooling system for a high current HTS DC reactor

    Science.gov (United States)

    Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

    2017-07-01

    A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

  18. Determination of the Design Speed of the Primary Cooling Pump in the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Chi, Daeyoung; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    An open-pool type research reactor is widely designed in consideration of the reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom. rimary cooling system circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the reactor core is continuously removed. After the primary cooling pumps stop, the decay heat is removed by the coastdown flow induced by the inertia force of a flywheel attached to each primary cooling pump. A pump coastdown flow means that the pump operates with the angular momentums of the shaft, impeller, and flywheel when a loss of electricity occurs. The primary cooling pump consists of the pump, flywheel, and moto. They are connected by flexible couplings. The primary cooling pump is conceptually designed based on the required flow rate and system constraints. A centrifugal pump of Case 1 with a non-dimensional specific speed of 0.59 and specific diameter of 4.94 is chosen as the primary cooling pump based on the hydraulic performance and mechanical integrity.

  19. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  20. Accurate Galactic Wind Simulations Require Gas Cooling to 10 K

    Science.gov (United States)

    Tanner, Ryan; Heitsch, Fabian; Cecil, Gerald N.

    2015-01-01

    Starbursts and AGN winds in galaxy cores can produce large-scale outflows or galactic winds (GW). Whether a starburst can form a GW depends on several variables including mechanical power into the ISM and the rate at which mass is loaded into the flow. Previous simulations (e.g. Hill+12, Cooper+08, Sutherland and Bicknell 2007) have included radiative cooling but only down to 10,000 K. We have modified the public Athena hydro code (Stone+08) to include a combined cooling curve from Sutherland and Dopita (1993) and Koyama and Inutsuka (2002) down to 10 K. We analyze grids of high-resolution 3D simulations of starbursts with an initial stellar mass ranging from 5e6 M⊙ to 1e8 M⊙. We find a 10-fold decrease of Hα emission in the halo resulting from the GW when we cool the gas down to 10 K vs the 10,000 K of previous simulations. We find that cooling to 10,000 K deposits 80% of the total GW gas mass in the warm phase (emitting Hα) whereas cooling to 10 K deposits only 7% in the warm phase but leaves 25% of the total GW gas mass in cold gas (cold temperatures, cold gas swept up into the halo by the GW is 4-5 orders of magnitude fainter than cold gas that remains in the disk. Thus detection of a cold GW component will be very difficult. Our results demonstrate that there are substantial differences in simulations with cooling down to 10 K vs cooling down to 10,000 K. Our work is funded by NASA/Herschel and NC Space Grant.

  1. Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

    2000-07-01

    The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the

  2. Helium turbine power generation in high temperature gas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuo [Tokyo Inst. of Tech. (Japan)

    1995-03-01

    This paper presents studies on the helium turbine power generator and important components in the indirect cycle of high temperature helium cooled reactor with multi-purpose use of exhaust thermal energy from the turbine. The features of this paper are, firstly the reliable estimation of adiabatic efficiencies of turbine and compressor, secondly the introduction of heat transfer enhancement by use of the surface radiative heat flux from the thin metal plates installed in the hot helium and between the heat transfer coil rows of IHX and RHX, thirdly the use of turbine exhaust heat to produce fresh water from seawater for domestic, agricultural and marine fields, forthly a proposal of plutonium oxide fuel without a slight possibility of diversion of plutonium for nuclear weapon production and finally the investigation of GT-HTGR of large output such as 500 MWe. The study of performance of GT-HTGR reduces the result that for the reactor of 450 MWt the optimum thermal efficiency is about 43% when the turbine expansion ratio is 3.9 for the turbine efficiency of 0.92 and compressor efficiency of 0.88 and the helium temperature at the compressor inlet is 45degC. The produced amount of fresh water is about 8640 ton/day. It is made clear that about 90% of the reactor thermal output is totally used for the electric power generation in the turbine and for the multi-puposed utilization of the heat from the turbine exhaust gas and compressed helium cooling seawater. The GT-Large HTGR is realized by the separation of the pressure and temperature boundaries of the pressure vessel, the increase of burning density of the fuel by 1.4 times, the extention of the nuclear core diameter and length by 1.2 times, respectively, and the enhancement of the heat flux along the nuclear fuel compact surface by 1.5 times by providing riblets with the peak in the flow direction. (J.P.N.).

  3. Low-power lead-cooled fast reactor for education purposes

    Directory of Open Access Journals (Sweden)

    D.S. Samokhin

    2015-11-01

    Full Text Available The possibility is examined to develop fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants. Main characteristics of liquid lead-cooled reactor using commercially implemented uranium dioxide with 19.7% enrichment with 235U isotope as the fuel load are examined. Hard neutron spectrum achieved in the fast reactor with compact core and natural lead coolant and, in longer term perspective, cooled with lead enriched with 208Pb isotope will allow addressing a number of research tasks under fast neutron flux densities of the order of 1013 neutrons/(cm2s. Relatively low thermal power equal to 0.5MW is envisaged for the purpose of safe handling of the reactor. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The studies are implemented based on the experience of development of low-power reactors available at the INPE NRNC “MEPhI”, as well as on the experience gained at the Joint-Stock Company “SSC RF-IPPE” in the field of development of fast reactors cooled with heavy liquid metal.

  4. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  5. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation.

  6. Estimation of turbulent mixing model for the application to liquid metal-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H. Y.; Ha, K. S.; Chang, W. P.; Lee, Y. B.; Heo, S

    2003-12-01

    It is required to model accurately the inter-subchannel mixing phenomenon for the improved prediction in the subchannel analysis and the flow blockage analysis of a Liquid Metal-cooled Reactor (LMR). When there exists a single-phase flow in the subchannels, the mixing of mass, energy and momentum between the subchannels can be divided into two parts, the diversion flow due to the pressure gradient and the cross flow mainly due to the turbulent mixing. To enlarge the understanding on turbulent mixing, the general turbulent models of zero-equation model, one-equation model and two-equation model are briefly introduced. Further, the turbulent mixing models, which are used in the subchannel codes such as MATRA-LMR, COBRA-IV, SABRE and ASFRE-III, are summarized. The bases of the turbulent mixing models in most subchannel codes are the mixing-length theory and the research results obtained before 1980's. The SABRE code includes the forms of one-equation model and two-equation model, but some experimental constants are essential to use those models. The recent experimental and analytical studies on turbulent mixing are surveyed and the important results are summarized. Some state-of-the-art turbulent mixing models are implemented in MATRA-LMR code and the effect of the models was evaluated for ORNL 19-pin data. The results imply the correlation by Rehme is the most suitable as a turbulent model for liquid metal-cooled reactors for wide range of fluidic conditions. To get more accurate distributions of flow and temperature for low flow conditions, it is recommended to have more accurate thermal conduction correction factor.

  7. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    Locke, B

    1998-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  8. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    LOCKE, B

    1999-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  9. Discussion on polonium extraction systems for Pb-PI-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buongiorno, J. [Idaho National Engineering and Environmental Lab., Nuclear Engineering Dept., Idaho Falls, ID (United States); Larson, C.L.; Czerwinski, K.R. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

    2001-07-01

    A discussion is presented on a polonium extraction technology that would reduce the radioactivity of the lead-bismuth coolant for fast reactors. This technology is based on the formation of the polonium hydride from the reaction of hydrogen gas with polonium-activated LBE. The equilibrium chemistry of the reaction was experimentally investigated. As a result, a correlation was generated for the free-energy of formation of the polonium hydride as a function of temperature. This correlation was then used for preliminary modeling of a polonium extraction system consisting in a mass exchanger where fine LBE droplets fall in countercurrent flow with a stream of pure hydrogen. It was found that a relatively compact and efficient polonium extraction system could be in principle designed, although significant technological and safety issues remain that are associated with the use and processing of hydrogen gas contaminated with polonium. (author)

  10. Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors

    Science.gov (United States)

    Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.

    2017-01-01

    Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..

  11. Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U.; Mueller, U. [Forschungszentrum Karlsruhe - Technik und Umwelt Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF), Karlsruhe (Germany)

    1997-12-31

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

  12. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    Science.gov (United States)

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  13. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  14. Experimental tests on the air cooling of the CLIC vertex detector

    CERN Document Server

    Duarte Ramos, Fernando; Nuiry, Francois-Xavier

    2016-01-01

    The strict requirements in terms of material budget for the inner region of the CLIC detector concept require the use of a dry gas for the cooling of the respective sensors. This, in conjunction with the compactness of the inner volumes, poses several challenges for the design of a cooling system that is able to fulfil the required detector specifications. This note summarizes the results obtained from experimental tests on the air cooling of the CLIC vertex detector as well as their comparison with the corresponding computational fluid dynamics simulations.

  15. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  16. Experimental and computational studies of film cooling with compound angle injection

    Energy Technology Data Exchange (ETDEWEB)

    Goldstein, R.J.; Eckert, E.R.G.; Patankar, S.V. [Univ. of Minnesota, Minneapolis, MN (United States)] [and others

    1995-10-01

    The thermal efficiency of gas turbine systems depends largely on the turbine inlet temperature. Recent decades have seen a steady rise in the inlet temperature and a resulting reduction in fuel consumption. At the same time, it has been necessary to employ intensive cooling of the hot components. Among various cooling methods, film cooling has become a standard method for cooling of the turbine airfoils and combustion chamber walls. The University of Minnesota program is a combined experimental and computational study of various film-cooling configurations. Whereas a large number of parameters influence film cooling processes, this research focuses on compound angle injection through a single row and through two rows of holes. Later work will investigate the values of contoured hole designs. An appreciation of the advantages of compound angle injection has risen recently with the demand for more effective cooling and with improved understanding of the flow; this project should continue to further this understanding. Approaches being applied include: (1) a new measurement system that extends the mass/heat transfer analogy to obtain both local film cooling and local mass (heat) transfer results in a single system, (2) direct measurement of three-dimensional turbulent transport in a highly-disturbed flow, (3) the use of compound angle and shaped holes to optimize film cooling performance, and (4) an exploration of anisotropy corrections to turbulence modeling of film cooling jets.

  17. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  18. Simulations of flow behavior of fuel particles in a conceptual helium-cooled spout fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang Shuyan; Li Xiang [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Lu Huilin [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)], E-mail: huilin@hit.edu.cn; Bouillard, Jacques [INERIS, Parc Technologique Alata, BP2, Verneuil-en-Halatte 60550 (France); Sun Qiaoqun; Wang Shuai [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)

    2009-01-15

    Hydrodynamics of helium and fuel particles are simulated in a conceptual helium-cooled spout fluidized bed nuclear reactor. The conceptual reactor consists of an axis-symmetric column with a sharp cone inside which the fuel particles are fluidized by helium. An isothermal gas-solid two-fluid flow model is presented. The kinetic-frictional constitutive model for dense assemblies of solids is incorporated. The kinetic stress is modeled using the kinetic theory of granular flow, while the friction stress is from the normal frictional stress model proposed by (Johnson, P.C., Nott, P., Jackson, R., 1990. Frictional-collisional equations of motion for particulate flows and their application to chutes. Journal of Fluid Mechanics 210, 501-535). Detailed spatial/temporal concentration and velocity profiles have been obtained in a conceptual spout fluidized bed nuclear reactor. The influence of inlet spouting jet velocity and conical angles on flow behavior of fluid and fuel particles is analyzed. The numerical simulations show that the unique mixing ability of the spout fluidized bed nuclear reactor gives rise, as expected, to uniform particle distributions. This uniformity enhances the heat transfer and therefore the power produced by the reactor.

  19. Versatile in situ gas analysis apparatus for nanomaterials reactors.

    Science.gov (United States)

    Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole

    2014-09-02

    We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (reactor, while suppressing the thermal decomposition of the analytes. It thus allows a more accurate study of the mechanism of progressive thermocatalytic cracking of precursors compared to previously reported conventional residual gas analyses of the reactor exhaust gas and hence paves the way for the controlled production of novel nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.

  20. Ducting arrangement for cooling a gas turbine structure

    Science.gov (United States)

    Lee, Ching-Pang; Morrison, Jay A.

    2015-07-21

    A ducting arrangement (10) for a can annular gas turbine engine, including: a duct (12, 14) disposed between a combustor (16) and a first row of turbine blades and defining a hot gas path (30) therein, the duct (12, 14) having raised geometric features (54) incorporated into an outer surface (80); and a flow sleeve (72) defining a cooling flow path (84) between an inner surface (78) of the flow sleeve (72) and the duct outer surface (80). After a cooling fluid (86) traverses a relatively upstream raised geometric feature (90), the inner surface (78) of the flow sleeve (72) is effective to direct the cooling fluid (86) toward a landing (94) separating the relatively upstream raised geometric feature (90) from a relatively downstream raised geometric feature (94).

  1. Entropy generation in a channel resembling gas turbine cooling ...

    Indian Academy of Sciences (India)

    Flow into a passage resembling a gas turbine blade cooling passage is considered and entropy generation rate in the passage is examined for unique rotation number and density ratios. In the simulations, leading and trailing walls of the passage are assumed to be at constant temperature. A control volume approach is ...

  2. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    Science.gov (United States)

    DeVolpi, A.

    1984-07-20

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  3. Measured gas and particle temperatures in VTT's entrained flow reactor

    DEFF Research Database (Denmark)

    Clausen, Sønnik; Sørensen, L.H.

    2006-01-01

    Particle and gas temperature measurements were carried out in experiments on VTTs entrained flow reactor with 5% and 10% oxygen using Fourier transform infrared emission spectroscopy (FTIR). Particle temperature measurements were performed on polish coal,bark, wood, straw particles, and bark...... and wood particles treated with additive. A two-color technique with subtraction of the background light was used to estimate particle temperatures during experiments. A transmission-emission technique was used tomeasure the gas temperature in the reactor tube. Gas temperature measurements were in good...

  4. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Administrator

    Cs, will be separated and used as a radiation source for various societal applications. This approach minimizes the quantity of waste to be immobilized. Separation of noble metals such as palladium for societal applications such as catalysts, fuel cells etc is also possible. 3. FBR Programme in India. The seed for fast reactor ...

  5. Solar coal gasification reactor with pyrolysis gas recycle

    Science.gov (United States)

    Aiman, William R.; Gregg, David W.

    1983-01-01

    Coal (or other carbonaceous matter, such as biomass) is converted into a duct gas that is substantially free from hydrocarbons. The coal is fed into a solar reactor (10), and solar energy (20) is directed into the reactor onto coal char, creating a gasification front (16) and a pyrolysis front (12). A gasification zone (32) is produced well above the coal level within the reactor. A pyrolysis zone (34) is produced immediately above the coal level. Steam (18), injected into the reactor adjacent to the gasification zone (32), reacts with char to generate product gases. Solar energy supplies the energy for the endothermic steam-char reaction. The hot product gases (38) flow from the gasification zone (32) to the pyrolysis zone (34) to generate hot char. Gases (38) are withdrawn from the pyrolysis zone (34) and reinjected into the region of the reactor adjacent the gasification zone (32). This eliminates hydrocarbons in the gas by steam reformation on the hot char. The product gas (14) is withdrawn from a region of the reactor between the gasification zone (32) and the pyrolysis zone (34). The product gas will be free of tar and other hydrocarbons, and thus be suitable for use in many processes.

  6. Recent results of research on supercritical water-cooled reactors in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Koehly, C.; Schulenberg, T. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Maraczy, C. [AEKI-KFKI, Budapest (Hungary); Toivonen, A.; Penttila, S. [VTT Technical Research Centre, Espoo (Finland); Chandra, L.; Lycklama a Nijeholt, J.A. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2009-07-01

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. The recent design of the HPLWR including flow paths is described in this paper. Exemplarily, design analyses are presented addressing neutronics, thermal-hydraulics, thermo-mechanics, materials investigations and heat transfer. (author)

  7. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  8. Natural gas turbine topping for the iris reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oriani, L.; Lombardi, C. [Politecnico di Milano, Milan (Italy); Paramonov, D. [Westinghouse Electric Corp., LLC, Pittsburgh, PA (United States)

    2001-07-01

    Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

  9. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  10. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  11. Flow analyses for the LAVA-ERVC experiment and the KSNP under the external reactor vessel cooling using RELAP5/MOD3 code

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik

    2005-01-01

    Flow analyses were performed using RELAP5/MOD3 code to investigate and verify the steam binding phenomena in the LAVA-ERVC experiment and to investigate the occurrence and the effects of steam binding for the KSNP under the external reactor vessel cooling. Flow analyses for the LAVA-ERVC experiments confirmed the steam binding occurrence in case of the limited steam venting and represented the LAVA-ERVC experimental results quite well. The flow analyses results for the KSNP under the external reactor vessel cooling address that water ingression and steam ventilation through the insulator are crucial factors determining the effective cool down via boiling heat removal at the outer surface of the RPV lower plenum. The flow analyses results for the base cases of the SBO and the 9.6 inch LBLOCA imply that the limited steam venting through the insulator induced the steam binding and eventually prevented the effective cooling at the outer surface of the RPV lower plenum. From the sensitivity study on the additional flow area for the steam venting, it could be found that the RPV lower plenum experienced effective cooling by smooth water circulation. The current RELAP5 flow analyses results for the KSNP under the external reactor vessel cooling address that prevention of steam binding phenomena should be settled first for the in-vessel corium retention through the external reactor vessel cooling. Implementation of additional flow path for the effective steam ventilation is highly recommended as one of the most promising countermeasures to enhance the coolability through the external reactor vessel cooling.

  12. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

    In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  13. Liquid-cooled nuclear reactor. Kernreaktor mit einem fluessigen Kuehlmittel

    Energy Technology Data Exchange (ETDEWEB)

    Deinlein, H.; Kummer, G.

    1984-07-26

    H/sub 2/ is directly added to the coolant circuit. This requires a pipe bypassing the volume expansion tank and being connected with the suction side of the high pressure pump. The supply of H/sub 2/ is realized via ceramic filter catridges in a liquid cooled part of the pipe at the suction side of the high pressure pump. Thus, the danger of oxyhydrogen explosions is avoided.

  14. Safe design of cooled tubular reactors for exothermic, multiple reactions. Consecutive reactions

    NARCIS (Netherlands)

    Westerterp, K.R.; Overtoom, R.R.M.

    1985-01-01

    The model of the pseudo-homogeneous, one-dimensional, cooled tubular reactor is applied to two consecutive, irreversible first order reactions. A criterion is derived to obtain a desired integral yield. Based on this criterion three requirements are formulated, which enable us to choose the relevant

  15. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more

  16. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    NARCIS (Netherlands)

    Steensma, M.; Westerterp, K.R.

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling

  17. A model to estimate volume change due to radiolytic gas bubbles and thermal expansion in solution reactors

    Energy Technology Data Exchange (ETDEWEB)

    Souto, F.J. [NIS-6: Advanced Nuclear Technology, Los Alamos National Lab., Los Alamos, NM (United States); Heger, A.S. [ESA-EA: Engineering Sciences and Application, Los Alamos National Lab., Los Alamos, NM (United States)

    2001-07-01

    To investigate the effects of radiolytic gas bubbles and thermal expansion on the steady-state operation of solution reactors at the power level required for the production of medical isotopes, a calculational model has been developed. To validate this model, including its principal hypotheses, specific experiments at the Los Alamos National Laboratory SHEBA uranyl fluoride solution reactor were conducted. The following sections describe radiolytic gas generation in solution reactors, the equations to estimate the fuel solution volume change due to radiolytic gas bubbles and thermal expansion, the experiments conducted at SHEBA, and the comparison of experimental results and model calculations. (author)

  18. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program: Topical report I, selection of candidate alloys. Volume 3. Selection of surface coating/substrate systems for screening creep and structural stability studies

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-20

    Considering the high temperature, low O/sub 2/, high C environment of operation in the Very High Temperature Reactor (VHTR) Systems, the utilization of coatings is envisaged to hold potential for extending component lifetimes through the formation of stable and continuous oxide films with enhanced resistance to C diffusion. A survey of the current state of technology for high temperature coatings has been performed. The usefulness of these coatings on the Mo, Ni, and Fe base alloys is discussed. Specifically, no coating substitute was identified for TZM other than the well known W-3 (pack silicide) and Al/sub 2/O/sub 3/ forming coatings were recommended for the Fe and Ni base structural materials. Recommendations as to coating types and processng have been made based on the predicted VHTR component size, shape, base metal and operational environment. Four tests designed to evaluate the effects of selected combinations of coatings and substrate matrices are recommended for consideration.

  19. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  20. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  1. Ambient air cooling arrangement having a pre-swirler for gas turbine engine blade cooling

    Science.gov (United States)

    Lee, Ching-Pang; Tham, Kok-Mun; Schroeder, Eric; Meeroff, Jamie; Miller, Jr., Samuel R; Marra, John J

    2015-01-06

    A gas turbine engine including: an ambient-air cooling circuit (10) having a cooling channel (26) disposed in a turbine blade (22) and in fluid communication with a source (12) of ambient air: and an pre-swirler (18), the pre-swirler having: an inner shroud (38); an outer shroud (56); and a plurality of guide vanes (42), each spanning from the inner shroud to the outer shroud. Circumferentially adjacent guide vanes (46, 48) define respective nozzles (44) there between. Forces created by a rotation of the turbine blade motivate ambient air through the cooling circuit. The pre-swirler is configured to impart swirl to ambient air drawn through the nozzles and to direct the swirled ambient air toward a base of the turbine blade. The end walls (50, 54) of the pre-swirler may be contoured.

  2. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  3. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—I: Development of criteria

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    Previously reported design criteria for cooled tubular reactors are based on the prevention of reactor temperature run away and were developed for single reactions only. In this paper it is argued that such criteri a should be based on the reactor selectivity, from which eventually a maximum

  4. The Lead Cooled Fast Reactor Benchmark BREST-300:. Analysis with Sensitivity Method

    Science.gov (United States)

    Smirnov, Valery; Orlov, Victor; Mourogov, Alexandre; Lecarpentier, David; Ivanova, Tatiana

    2006-04-01

    Sustainable development of atomic energy will require development of new types of reactors able to exceed the limits of the existing reactor types in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, economic and safety competitiveness. Lead cooled fast neutrons reactor is one of the most interesting candidates with a potential to address these needs. BREST-300 is a 300 MWe lead cooled fast reactor developed by the NIKIET (Russia) with a deterministic safety approach which aims to exclude reactivity margins greater than the delayed neutron fraction. The development of innovative reactors (lead coolant, nitride fuel…) and fuel cycles with new constraints such as cycle closure or actinide burning, requires new technologies and new data from various disciplines: fuel types, fuel designs and fuel reprocessing. In this connection, the tool and neutron data used for the calculational analysis of reactor characteristics requires thorough validation, even if computational codes in Russia and France relies to the calculation of fast reactors' parameters and “fast” experiments. NIKIET developed a reactor benchmark fitting of design type calculational tools (including neutron data). In the frame of technical exchanges between the NIKIET and the EDF (Electricité De France), results of this benchmark calculation concerning the principal parameters of fuel evolution and safety parameters has been intercompared, in order to estimate the uncertainties and validate the codes for calculations of these new kind of reactors. Different codes and cross-sections data have been used, and sensitivity studies have been performed to understand and quantify the uncertainties sources.

  5. Gas-phase photocatalysis in μ-reactors

    DEFF Research Database (Denmark)

    Vesborg, Peter Christian Kjærgaard; Olsen, Jakob Lind; Henriksen, Toke Riishøj

    2010-01-01

    Gas-phase photocatalysis experiments may benefit from the high sensitivity and good time response in product detection offered by μ-reactors. We demonstrate this by carrying out CO oxidation and methanol oxidation over commercial TiO2 photocatalysts in our recently developed high-sensitivity reac......Gas-phase photocatalysis experiments may benefit from the high sensitivity and good time response in product detection offered by μ-reactors. We demonstrate this by carrying out CO oxidation and methanol oxidation over commercial TiO2 photocatalysts in our recently developed high...

  6. Gas turbine cooling modeling - Thermodynamic analysis and cycle simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jordal, Kristin

    1999-02-01

    Considering that blade and vane cooling are a vital point in the studies of modern gas turbines, there are many ways to include cooling in gas turbine models. Thermodynamic methods for doing this are reviewed in this report, and, based on some of these methods, a number of model requirements are set up and a Cooled Gas Turbine Model (CGTM) for design-point calculations of cooled gas turbines is established. Thereafter, it is shown that it is possible to simulate existing gas turbines with the CGTM. Knowledge of at least one temperature in the hot part of the turbine (TET, TRIT or possibly TIT) is found to be vital for a complete heat balance over the turbine. The losses, which are caused by the mixing of coolant and main flow, are in the CGTM considered through a polytropic efficiency reduction factor S. Through the study of S, it can be demonstrated that there is more to gain from coolant reduction in a small and/or old turbine with poor aerodynamics, than there is to gain in a large, modern turbine, where the losses due to interaction between coolant and main flow are, relatively speaking, small. It is demonstrated, at the design point (TET=1360 deg C, {pi}=20) for the simple-cycle gas turbine, that heat exchanging between coolant and fuel proves to have a large positive impact on cycle efficiency, with an increase of 0.9 percentage points if all of the coolant passes through the heat exchanger. The corresponding improvement for humidified coolant is 0.8 percentage points. A design-point study for the HAT cycle shows that if all of the coolant is extracted after the humidification tower, there is a decrease in coolant requirements of 7.16 percentage points, from 19.58% to 12.52% of the compressed air, and an increase in thermal efficiency of 0.46 percentage points, from 53.46% to 53.92%. Furthermore, it is demonstrated with a TET-parameter variation, that the cooling of a simple-cycle gas turbine with humid air can have a positive effect on thermal efficiency

  7. Flue gas injection control of silica in cooling towers.

    Energy Technology Data Exchange (ETDEWEB)

    Brady, Patrick Vane; Anderson, Howard L., Jr.; Altman, Susan Jeanne

    2011-06-01

    Injection of CO{sub 2}-laden flue gas can decrease the potential for silica and calcite scale formation in cooling tower blowdown by lowering solution pH to decrease equilibrium calcite solubility and kinetic rates of silica polymerization. Flue gas injection might best inhibit scale formation in power plant cooling towers that use impaired makeup waters - for example, groundwaters that contain relatively high levels of calcium, alkalinity, and silica. Groundwaters brought to the surface for cooling will degas CO{sub 2} and increase their pH by 1-2 units, possibly precipitating calcite in the process. Recarbonation with flue gas can lower the pHs of these fluids back to roughly their initial pH. Flue gas carbonation probably cannot lower pHs to much below pH 6 because the pHs of impaired waters, once outgassed at the surface, are likely to be relatively alkaline. Silica polymerization to form scale occurs most rapidly at pH {approx} 8.3 at 25 C; polymerization is slower at higher and lower pH. pH 7 fluids containing {approx}220 ppm SiO{sub 2} require > 180 hours equilibration to begin forming scale whereas at pH 8.3 scale formation is complete within 36 hours. Flue gas injection that lowers pHs to {approx} 7 should allow substantially higher concentration factors. Periodic cycling to lower recoveries - hence lower silica concentrations - might be required though. Higher concentration factors enabled by flue gas injection should decrease concentrate volumes and disposal costs by roughly half.

  8. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David [Idaho National Lab. (INL), Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab. (INEEL); Martin, Philippe [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Phelip, Mayeul [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Ballinger, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2004-12-01

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  9. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  10. The Formation and Physical Origin of Highly Ionized Cooling Gas

    Science.gov (United States)

    Bordoloi, Rongmon; Wagner, Alexander Y.; Heckman, Timothy M.; Norman, Colin A.

    2017-10-01

    We present a simple model that explains the origin of warm, diffuse gas seen primarily as highly ionized absorption-line systems in the spectra of background sources. We predict the observed column densities of several highly ionized transitions such as O vi, O vii, Ne viii, N v, and Mg x, and we present a unified comparison of the model predictions with absorption lines seen in the Milky Way disk, Milky Way halo, starburst galaxies, the circumgalactic medium, and the intergalactic medium at low and high redshifts. We show that diffuse gas seen in such diverse environments can be simultaneously explained by a simple model of radiatively cooling gas. We show that most such absorption-line systems are consistent with being collisionally ionized, and we estimate the maximum-likelihood temperature of the gas in each observation. This model satisfactorily explains why O vi is regularly observed around star-forming low-z L* galaxies, and why N v is rarely seen around the same galaxies. We further present some consequences of this model in quantifying the dynamics of the cooling gas around galaxies and predict the shock velocities associated with such flows. A unique strength of this model is that while it has only one free (but physically well-constrained) parameter, it nevertheless successfully reproduces the available data on O vi absorbers in the interstellar, circumgalactic, intragroup, and intergalactic media, as well as the available data on other absorption lines from highly ionized species.

  11. Risk Based Inspection of Gas-Cooling Heat Exchanger

    Directory of Open Access Journals (Sweden)

    Dwi Priyanta

    2017-09-01

    Full Text Available On October 2013, Pertamina Hulu Energi Offshore North West Java (PHE – ONWJ platform personnel found 93 leaking tubes locations in the finfan coolers/ gas-cooling heat exchanger. After analysis had been performed, the crack in the tube strongly indicate that stress corrosion cracking was occurred by chloride. Chloride stress corrosion cracking (CLSCC is the cracking occurred by the combined influence of tensile stress and a corrosive environment. CLSCC is the one of the most common reasons why austenitic stainless steel pipework or tube and vessels deteriorate in the chemical processing, petrochemical industries and maritime industries. In this thesis purpose to determine the appropriate inspection planning for two main items (tubes and header box in the gas-cooling heat exchanger using risk based inspection (RBI method. The result, inspection of the tubes must be performed on July 6, 2024 and for the header box inspection must be performed on July 6, 2025. In the end, RBI method can be applicated to gas-cooling heat exchanger. Because, risk on the tubes can be reduced from 4.537 m2/year to 0.453 m2/year. And inspection planning for header box can be reduced from 4.528 m2/year to 0.563 m2/year.

  12. Numerical and Experimental Study of a Cooling for Vanes in a Small Turbine Engine

    Science.gov (United States)

    Šimák, Jan; Michálek, Jan

    2016-03-01

    This paper is concerned with a cooling system for inlet guide vanes of a small turbine engine which are exposed to a high temperature gas leaving a combustion chamber. Because of small dimensions of the vanes, only a simple internal cavity and cooling holes can be realized. The idea was to utilize a film cooling technique. The proposed solution was simulated by means of a numerical method based on a coupling of CFD and heat transfer solvers. The numerical results of various scenarios (different coolant temperature, heat transfer to surroundings) showed a desired decrease of the temperature, especially on the most critical part - the trailing edge. The numerical data are compared to results obtained by experimental measurements performed in a test facility in our institute. A quarter segment model of the inlet guide vanes wheel was equipped with thermocouples in order to verify an effect of cooling. Despite some uncertainty in the results, a verifiable decrease of the vane temperature was observed.

  13. Modelling of non-catalytic reactors in a gas-solid trickle flow reactor: Dry, regenerative flue gas desulphurization using a silica-supported copper oxide sorbent

    NARCIS (Netherlands)

    Kiel, J.H.A.; Kiel, J.H.A.; Prins, W.; van Swaaij, Willibrordus Petrus Maria

    1992-01-01

    A one-dimensional, two-phase dispersed plug flow model has been developed to describe the steady-state performance of a relatively new type of reactor, the gas-solid trickle flow reactor (GSTFR). In this reactor, an upward-flowing gas phase is contacted with as downward-flowing dilute solids phase

  14. Analysis of the optimal fuel composition for the Indonesian experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.), Ibaraki (Japan); Sembiring, Tagor Malem [National Nuclear Energy Agency of Indonesia, Banten (Indonesia). Center for Nuclear Reactor Technology and Safety; Arbie, Bakri; Subki, Iyos [PT MOTAB Technology, Jakarta Barat (Indonesia)

    2017-03-15

    The optimal fuel composition of the 10 MWth Experimental Power Reactor (RDE), to be built by the Indonesian National Nuclear Energy Agency (BATAN), is a very important design parameter since it will directly affect the fuel cost, new and spent fuel storage capacity, and other back-end environmental burden. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO{sub 2} TRISO fuel under multipass or once-through-then-out fueling scheme. A scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment is conducted. All burnup, criticality calculations and core equilibrium search are carried out by using BATAN-MPASS, a general in-core fuel management code for pebble bed HTGRs, featured with many automatic equilibrium searching options as well as thermal-hydraulic calculation capability. The RDE User Requirement Document issued by BATAN is used to derive the main core design parameters and constraints. The scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4 g to 10 g/pebble. Fissile loading per unit energy generated (kg/GWd) is taken as the objective function for the present scoping study. The analysis results show that the optimal HM loading is around 8 g/pebble. Under the constraint of 80 GWd/t fuel discharge burnup imposed by the technical specification, the uranium enrichment for the optimal HM loading is approximately 13 w/o.

  15. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  16. Theoretical thermodynamics analysis of cooling cycle bu advanced gas absorption using solar energy; Analisis teorico-experimental de un ciclo de refrigeracion por absorcion avanzado gax, operando con energia solar

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, V. E.; Vidal, A. S.; Garcia, C. A.; Garcia-Valladares, O.; Best, R. B.; Hernandez, J. G.; Velazquez, N. L.

    2004-07-01

    In this article a solar system of refrigeration by absorption with heat exchange generator absorber (GAX) was analyzed. A theoretical thermodynamic analysis of the energetic behavior of the GAX absorption system was made. Experimental results were obtained with generation temperatures of 190 and 220 C, the evaporation temperature was set at 9 C and temperatures of cooling fluids (air and water) were set at 30 C and 28 C, respectively. It was possible to appreciate that the GAX effect decrease whether absorber, type falling film, is operated in option of parallel flow and it was increased when the absorber was operated in option of counterflow. (Author)

  17. Flow and heat transfer investigations in swirl tubes for gas turbine blade cooling

    OpenAIRE

    Biegger, Christoph

    2017-01-01

    A swirl tube is a very effective cooling technique for high thermal loaded components like gas turbine blades. Such a tube consists of one or more tangential inlet jets, which induce a highly 3D swirling flow. This swirling flow is characterized by large velocities near the wall and an enhanced turbulence in the tube which both increase the convective heat transfer. In the present work, the flow phenomena and the heat transfer in swirl tubes are studied experimentally and numerically. Therefo...

  18. Adaptation of Phytoplankton-Degrading Microbial Communities to Thermal Reactor Effluent in a New Cooling Reservoir

    Science.gov (United States)

    Schoenberg, Steven A.; Benner, Ronald; Sobecky, Patricia; Hodson, Robert E.

    1988-01-01

    In water column and sediment inocula from a nuclear reactor cooling reservoir, natural phytoplankton substrate labeled with 14C was used to determine aerobic and anaerobic mineralization rates for a range of temperatures (25, 40, 55, and 70°C) expected during reactor operation. For experiments that were begun during reactor shutdown, aerobic decomposition occurred at temperatures of <55°C. After 2 months of reactor operation, aerobic rates increased substantially at 55 and 70°C, although maximum rates were observed at temperatures of ≤40°C. The temperature range for which maximum anaerobic mineralization (i.e., the sum of CH4 and CO2) was observed was 25 to 40°C when the reactor was off, expanding to 25 to 55°C during reactor operation. Increased rates at 55°C, but not 70°C, correlated with an increase in the ratio of cumulative methane to carbon dioxide produced over 21 days. When reduced reactor power lowered the maximum temperature of the reservoir to 42°C, aerobic decomposition at 70°C was negligible, but remained substantial at 55°C. Selection for thermophilic decomposers occurred rapidly in this system in both aerobic and anaerobic communities and did not require prolonged exposure to elevated temperatures. PMID:16347659

  19. Gas and liquid distribution in the monolith film flow reactor

    NARCIS (Netherlands)

    Heibel, A.K.; Vergeldt, F.J.; As, van H.

    2003-01-01

    The gas-liquid distribution in a monolith film flow reactor is investigated in the scope of this work. Magnetic resonance imaging (MRI) and a customized liquid collection method hate been successfully applied to determine the liquid distribution over the monolith cross-section. Using a

  20. CFD study on the supercritical carbon dioxide cooled pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dali, E-mail: ydlmitd@outlook.com; Peng, Minjun; Wang, Zhongyi

    2015-01-15

    Highlights: • An innovation concept of supercritical carbon dioxide cooled pebble bed reactor is proposed. • Body-centered cuboid (BCCa) arrangement is adopted for the pebbles. • S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor. - Abstract: The thermal hydraulic study of using supercritical carbon dioxide (S-CO{sub 2}), a superior fluid state brayton cycle medium, in pebble bed type nuclear reactor is assessed through computational fluid dynamics (CFD) methodology. Preliminary concept design of this S-CO{sub 2} cooled pebble bed reactor (PBR) is implemented by the well-known KTA heat transfer correlation and Ergun pressure drop equation. Eddy viscosity transport turbulence model is adopted and verified by KTA calculated results. Distributions of the temperature, velocity, pressure and Nusselt (Nu) number of the coolant near the surface of the middle spherical fuel element are obtained and analyzed. The conclusion of the assessment is that S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor due primarily to its good heat transfer characteristic and large mass density, which could lead to achieve lower pressure drop and higher power density.

  1. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  2. Afterheat removal from a helium reactor under accident conditions. CFD calculations for the code-to-code benchmark analyses on the thermal behavior for the gas turbine modular helium reactor

    Energy Technology Data Exchange (ETDEWEB)

    Siccama, N.B.; Koning, H

    1998-04-01

    The International Atomic Energy Agency (IAEA) Co-ordinated Research Programme (CRP) on `Heat Transport and Afterheat Removal for Gas Cooled Reactors under Accident Conditions` has organised benchmark analyses to support verification and validation of analytical tools used by the participants to predict the thermal behaviour of advanced gas cooled reactors during accidents. One of thew benchmark analyses concerns the code-to-code analysis of the Gas Turbine Modular Helium Reactor (GT-MHR) plutonium burner accidents. The GT-MHR is a passive safe, helium cooled, graphite moderated, advanced reactor system with a thermal power of 600 MW that is based on existing technology. The GT-MHR can also be fuelled with plutonium. If the main helium cooling and the auxiliary shut-down cooling systems fail or become unavailable, the core afterheat is removed by radiation and convection inside the reactor vessel and the reactor cavity to the Reactor Cavity Cooling System (RCCS). The objective of the RCCS is to serve as an ultimate heat sink, ensuring the thermal integrity of the core, vessel and critical equipment within the reactor cavity for the entire spectrum of postulated accident sequences. This paper describes the heat transport inside the reactor core to the RCCS. For this purpose, the heat transfer mechanisms as well as the flow patterns inside the core, the reactor pressure vessel, and the cavity have been calculated by the Computational Fluid Dynamics (CFD) code CFX-F3D. The behaviour of the RCCS itself is not described. One calculation considers the full power operation, while two calculations consider Loss Of Forced Convection (LOFC) accidents, one at pressurised conditions and the other depressurised conditions. The heat transfer from the reactor vessel to the environment under normal operation conditions is 2.64 MW. The highest temperature in the core is 1222K, and the average core temperature is 1075K. The highest reactor vessel temperature is 679K. The highest

  3. Drop Performance Test of Conceptually Designed Control Rod Assembly for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Young-Kyu Lee

    2017-06-01

    Full Text Available The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  4. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  5. Breeding zone models of DEMO ceramic helium cooled blanket test module for testing in IVV-2M reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kovalenko, V.; Kiryiak, L.; Lopatkin, A.; Marachev, A.; Muratov, V.; Strebkov, Yr. [Federal State Unitary Enterprise ' ' Dollezhal Research and Development Inst. of Power Engineering' ' , Moscow (Russian Federation); Davydov, D.; Kapyshev, V.; Kazennov, Yr.; Tebus, V. [Federal State Unitary Enterprise ' ' A.A. Bochvar All-Russia Research Inst. of Inorganic Materials' ' , Moscow (Russian Federation)

    2002-06-01

    The goal of DEMO ceramic helium cooled blanket test module (CHC BTM) is to demonstrate a breeding capability that would lead to tritium self-sufficiency in ITER reactor and to extract a high-grade heat suitable for electricity generation. Experimental validation of all the adopted design solutions is main important problem at design and calculation works carrying out in order to develop the CHC BTM. One important task for breeding zones feasibility validation is in-pile tests. Two models were developed and fabricated for testing in the fission IVV-2M reactor. Breeding zone is based on poloidal BIT-conception. The models structural material is ferrito-martensitic steel. Breeder material is lithium orthosilicate in pebble beds and pellet forms. Multiplier material is beryllium in pebble beds and porosity forms. The cooling is provided by helium at 10 MPa. The tritium produced in the breeder material is purged by the helium flow at 0.1-0.2 MPa. Designs of model description and experimental channel, results of neutronic and thermo-hydraulic calculations are presented in the paper. (orig.)

  6. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  7. Improvements on cool gas generators and their application in space propulsion systems

    NARCIS (Netherlands)

    Sanders, H.M.; Schuurbiers, C.A.H.; Vandeberg, R.J.

    2014-01-01

    Cool Gas Generators are an innovative means to store gas which can be used in propulsion and pressurization systems but also for inflatable structures and terrestrial applications. In Cool Gas Generators, the gas is stored chemically, without pressure or leakage and with a long life time without

  8. The Status of the US High-Temperature Gas Reactors

    Directory of Open Access Journals (Sweden)

    Andrew C. Kadak

    2016-03-01

    Full Text Available In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Congress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a “hydrogen” economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

  9. Draft layout, containment and performance of the safety system of the European Supercritical Water-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schlagenhaufer, M.; Kohly, C.; Schulenberg, T. [Karlsruhe Inst. of Tech., Karlsruhe (Germany); Rothschmitt, S.; Bittermann, D. [AREVA NP GmbH, Erlangen (Germany)

    2010-07-01

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. Close to the end of the project the technical results are translated into a draft layout of the HPLWR. The containment and safety system are being explained. Exemplarily, a depressurization event shows the capabilities of the safety system to sufficiently cool the reactor by means of a low pressure coolant injection system. (author)

  10. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.

    1976-01-01

    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  11. Upgrading program of the experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, A.; Yogo, S. [Japan Nuclear Cycle Development Institute, Iibaraki-Ken (Japan)

    2001-07-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  12. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  13. Phytoplankton distribution in three thermally different but edaphically similar reactor cooling reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Wilde, E W

    1982-01-01

    Phytoplankton community structure and the physicochemical characteristics of three reactor cooling reservoirs in close proximity and of similar age and bottom type were studied during 1978. The three reservoirs differed in thermal alteration resulting from reactor cooling water as follows: (1) considerable heating with lake-wide temperatures >30/sup 0/C, even in winter; (2) a maximal 5/sup 0/C increase occurring in only one of three major arms of the reservoir; and (3) no thermal effluent received during the study period. Considerable spatial and temporal differences in water quality and phytoplankton community structure were observed; however, water temperature independent of other environmental factors (e.g., light and nutrients) was found to be a relatively unimportant variable for explaining phytoplankton periodicity.

  14. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  15. The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

    Energy Technology Data Exchange (ETDEWEB)

    Silberstein, A.J. [Framatome, Paris (France)

    1998-09-01

    As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

  16. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    Science.gov (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  17. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  18. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  19. The Small Modular Liquid Metal Cooled Reactor: A New Approach to Proliferation Risk Management

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C F; Crawford, D; Cappiello, M; Minato, A; Herczeg, J W

    2003-11-12

    There is an ongoing need to supply energy to small markets and remote locations with limited fossil fuel infrastructures. The Small, Modular, Liquid-Metal-Cooled Reactor, also referred to as SSTAR (Small, Secure, Transportable, Autonomous Reactor), can provide reliable and cost-effective electricity, heat, fresh water, and potentially hydrogen transportation fuels for these markets. An evaluation of a variety of reactor designs indicates that SSTAR, with its secure, long-life core, has many advantages for deployment into a variety of national and international markets. In this paper, we describe the SSTAR concept and its approach to safety, security, environmental and non-proliferation. The system would be design-certified using a new license-by-test approach, and demonstrated for commercial deployment anywhere in the world. The project addresses a technology development need (i.e., a small secure modular system for remote sites) that is not otherwise addressed in other currently planned research programs.

  20. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices.

  1. Innovative radiation-based direct heat exchanger (DHX) for liquid metal cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    De Santis, Andrea, E-mail: andrea.desantis@uniroma1.it [“SAPIENZA” University of Rome, DIAEE, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Vitale Di Maio, Damiano; Caruso, Gianfranco [“SAPIENZA” University of Rome, DIAEE, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Manni, Fabio [S.R.S. Servizi di Ricerche e Sviluppo S.r.l., Rome (Italy)

    2013-10-15

    Highlights: • An innovative DHRS for liquid metal fast breeder reactors has been proposed. • A parametric CFD analyses of the DHX performances have been performed. • A comparison between SFR and LFR applications has been performed. -- Abstract: Considering the importance of safety features in the development of Generation IV nuclear reactors, an innovative and passive decay heat removal system (DHRS) has been proposed for liquid metal cooled reactors. The attention is here focused on the direct heat exchanger (DHX) of the system constituted by a bayonet tube that allows to remove the decay heat from the primary coolant; both primary and secondary fluids flow in natural circulation. Since each bayonet tube is equipped with a vacuum gap, the most important heat transfer mechanism characterizing the DHX is radiation. Furthermore, the presence of the vacuum gap guarantees a physical separation and a complete decoupling between primary and secondary fluids, enhancing the safety features of the whole system. Several CFD analyses have been carried out in order to obtain a characterization of the DHX both for sodium and lead cooled fast reactors, in order to optimize the DHX geometry on the basis of the specific application, and the results are discussed in the paper.

  2. Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

    National Research Council Canada - National Science Library

    Zhang, Guo-Qing; Wang, Shuai; Zhang, Hai-Qing; Zhu, Xing-Wang; Peng, Chao; Cai, Jun; He, Zhao-Zhong; Chen, Kun

    2017-01-01

    The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR...

  3. Nuclear power station with a water-cooled reactor pressure vessel. Kernkraftwerk mit einem wassergekuehlten Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, R.; Brunner, G.; Jost, N.

    1987-10-29

    Nuclear radiation produces radiolysis gases, which are undesirable for corrosion and oxyhydrogen gas reasons. To limit the proportion of this radiolysis gas, the invention provides that catalytic surfaces should be introduced into the primary circuit, to produce recombination of hydrogen and oxygen. These surfaces can be accommodated in the upper part of the reactor pressure vessel. The live steam screen can also have a catalytic surface.

  4. The experimental study of neutralized electron beams for electron cooling

    CERN Document Server

    Bosser, Jacques; MacCaferri, R; Molinari, G; Tranquille, G; Varenne, F; Korotaev, Yu V; Meshkov, I N; Polyakov, V A; Smirnov, A; Syresin, E M

    1996-01-01

    In this report we present the latest experimental results on electron beam neutralization. These experiments have been made at LEAR and on the JINR test bench. The main difficulty in obtaining neutralized beams resides in an instability which is dependent on the electron beam current. A number of methods have been developed in order to overcome this instability and have enabled us to further investigate the possibility of generating intense low energy electron beams for the cooling of Pb ions.

  5. Experimental and numerical investigation of gas phase freeboard combustion

    DEFF Research Database (Denmark)

    Andersen, J.; Jensen, Peter Arendt; Meyer, K.E.

    2009-01-01

    Experimental data for velocity field, temperatures, and gas composition have been obtained from a 50 kW axisymmetric non-swirling natural gas fired combustion setup under two different settings. The reactor was constructed to simulate the conditions in the freeboard of a grate-fired boiler but un...... of more advanced chemical mechanisms did not improve the prediction of the overall combustion process but did provide additional information about species (especially H(2) and radicals), which is desirable for postprocessing pollutant formation.......Experimental data for velocity field, temperatures, and gas composition have been obtained from a 50 kW axisymmetric non-swirling natural gas fired combustion setup under two different settings. The reactor was constructed to simulate the conditions in the freeboard of a grate-fired boiler...... but under well-defined conditions. The experimental results are compared to computational fluid dynamics (CFD) modeling predictions, using the eddy dissipation model (EDM) its well as the eddy dissipation concept (EDC). The use of EDC allows for implementation of more advanced combustion schemes; we have...

  6. Breeding gains of sodium-cooled oxide-fueled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mougniot, J. C.; Barre, J. Y.; Clauzon, P.; Ciacometti, C.; Neviere, G.; Ravier, J.; Sichard, B.

    1975-12-01

    Calculated values are presented for the breeding gains of French fast reactors, and the experimental uncertainties are discussed. The effect of various choices of planning on the breeding gains is next analyzed within the framework of classical concepts. In the final part, a new concept involving heterogeneous cores with a single enrichment zone is presented. This concept permits a significant improvement in the breeding gain and doubling time of fast reactors.

  7. Computational and experimental study on supersonic film cooling for liquid rocket nozzle applications

    Directory of Open Access Journals (Sweden)

    Vijayakumar Vishnu

    2015-01-01

    Full Text Available An experimental and computational investigation of supersonic film cooling (SFC was conducted on a subscale model of a rocket engine nozzle. A computational model of a convergent-divergent nozzle was generated, incorporating a secondary injection module for film cooling in the divergent section. Computational Fluid Dynamic (CFD simulations were run on the model and different injection configurations were analyzed. The CFD simulations also analyzed the parameters that influence film cooling effectiveness. Subsequent to the CFD analysis and literature survey an angled injection configuration was found to be more effective, therefore the hardware was fabricated for the same. The fabricated nozzle was later fixed to an Air-Kerosene combustor and numerous sets of experiments were conducted in order to ascertain the effect on film cooling on the nozzle wall. The film coolant employed was gaseous Nitrogen. The results showed substantial cooling along the walls and a considerable reduction in heat transfer from the combustion gas to the wall of the nozzle. Finally the computational model was validated using the experimental results. There was fairly good agreement between the predicted nozzle wall temperature and the value obtained through experiments.

  8. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  9. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  10. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  11. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL; Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Holcomb, David Eugene [ORNL

    2011-01-01

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

  12. Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor

    Science.gov (United States)

    Arda, Samet Egemen

    A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.

  13. Advanced Water-Gas Shift Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

    2009-01-07

    The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

  14. Experimental investigation on ejecting low-temperature cooling superconducting magnets

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Bin; Zhang, Qiang, E-mail: 6266798@qq.com; Tong, Ming-wei; Hu, Peng; Wu, Shuang-ying; Cai, Qin; Qin, Zeng-hu

    2013-10-15

    Highlights: • The cooling temperature of the superconducting materials can be adjusted by the ejecting refrigeration. • The result shows that the temperature of liquid nitrogen can be reduced to 70 K by controlling the inlet water pressure of the ejector. • The refrigeration performance of ejector is affected by the different structure and system pressure. -- Abstract: With the development of the high-temperature superconducting (HTS) materials and refrigeration technologies, using ejecting refrigeration to cool the superconducting materials becomes the direction of HTS applications. In this paper, an experimental study has been carried out on the basis of the theory of analyzing the ejecting low-temperature cooling superconducting magnet. The relationship between area ratios and refrigeration performance at different system pressures was derived. In addition, the working fluid flow and suction chamber pressure of the ejector with different area ratios at various inlet pressures have been examined to obtain the performance of ejectors under different working conditions. The result shows that the temperature of liquid nitrogen can be reduced to 70 K by controlling the inlet water pressure when the pressurized water at 20 °C is used to eject the saturated liquid nitrogen, which can provide the stable operational conditions for the HTS magnets cooling.

  15. Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A., E-mail: abanades@etsii.upm.es [ETSII/Universidad Politecnica de Madrid, J.Gutierrez Abascal, 2-28006 Madrid (Spain); Garcia, C.; Garcia, L. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba); Escriva, A.; Perez-Navarro, A. [Instituto de Ingenieria Energetica, Universidad Politecnica de Valencia, C.P. 46022 Valencia (Spain); Rosales, J. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba)

    2011-06-15

    Highlights: > Utilization of Accelerator Driven System (ADS) for Hydrogen production. > Evaluation of the potential use of gas-cooled ADS for a sustainable use of Uranium resources by transmutation of nuclear wastes, electricity and Hydrogen production. > Application of the Sulfur-Iodine thermochemical process to subcritical systems. > Application of CINDER90 to calculate burn-up in subcritical systems. - Abstract: The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45-46%, either in the form of Hydrogen, electricity, or both.

  16. Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

    Science.gov (United States)

    Abedi-Varaki, Mehdi

    2017-08-01

    Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

  17. An integration scheme for stiff solid-gas reactor models

    Directory of Open Access Journals (Sweden)

    Bjarne A. Foss

    2001-04-01

    Full Text Available Many dynamic models encounter numerical integration problems because of a large span in the dynamic modes. In this paper we develop a numerical integration scheme for systems that include a gas phase, and solid and liquid phases, such as a gas-solid reactor. The method is based on neglecting fast dynamic modes and exploiting the structure of the algebraic equations. The integration method is suitable for a large class of industrially relevant systems. The methodology has proven remarkably efficient. It has in practice performed excellent and been a key factor for the success of the industrial simulator for electrochemical furnaces for ferro-alloy production.

  18. Drop Performance Test of Conceptually Designed Control Rod Assembly for Prototype Generation IV Sodium-cooled Fast Reactor

    OpenAIRE

    Lee, Young-Kyu; Lee, Jae-Han; Kim, Hoe-Woong; Kim, Sung-Kyun; Kim, Jong-Bum

    2017-01-01

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as...

  19. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    Science.gov (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  20. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  1. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  2. Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

    Science.gov (United States)

    Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

    2017-12-01

    A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

  3. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  4. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    Science.gov (United States)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

  5. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  6. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  7. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40 %) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2–3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  8. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  9. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  10. Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

    Science.gov (United States)

    Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

    2017-08-01

    The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

  11. Investigation of the Thermal Performance of a Vertical Two-Phase Closed Thermosyphon as a Passive Cooling System for a Nuclear Reactor Spent Fuel Storage Pool

    Directory of Open Access Journals (Sweden)

    Mukhsinun Hadi Kusuma

    2017-04-01

    Full Text Available The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  12. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  13. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  14. A new gas cooling model for semi-analytic galaxy formation models

    Science.gov (United States)

    Hou, Jun; Lacey, Cedric G.; Frenk, Carlos S.

    2018-03-01

    Semi-analytic galaxy formation models are widely used to gain insight into the astrophysics of galaxy formation and in model testing, parameter space searching and mock catalogue building. In this work, we present a new model for gas cooling in haloes in semi-analytic models, which improves over previous cooling models in several ways. Our new treatment explicitly includes the evolution of the density profile of the hot gas driven by the growth of the dark matter halo and by the dynamical adjustment of the gaseous corona as gas cools down. The effect of the past cooling history on the current mass cooling rate is calculated more accurately, by doing an integral over the past history. The evolution of the hot gas angular momentum profile is explicitly followed, leading to a self-consistent and more detailed calculation of the angular momentum of the cooled down gas. This model predicts higher cooled down masses than the cooling models previously used in GALFORM, closer to the predictions of the cooling models in L-GALAXIES and MORGANA, even though those models are formulated differently. It also predicts cooled down angular momenta that are higher than in previous GALFORM cooling models, but generally lower than the predictions of L-GALAXIES and MORGANA. When used in a full galaxy formation model, this cooling model improves the predictions for early-type galaxy sizes in GALFORM.

  15. Design of a supercritical water-cooled reactor with a three-pass core arrangement

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, K. [EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg, D-76661 Philippsburg (Germany)], E-mail: kai-fischer@gmx.de; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany); Laurien, E. [University of Stuttgart, Institute for Nuclear and Energy Systems (IKE), Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2009-04-15

    The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected. Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 deg. C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the reactor.

  16. Simulation of Water Gas Shift Zeolite Membrane Reactor

    Science.gov (United States)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  17. Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

  18. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  19. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  20. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  1. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, Paul R.

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  2. Comprehensive Prediction of Thermosyphon Characteristics in Reactor Passive Cooling System Simulation Loop FASSIP-01

    Directory of Open Access Journals (Sweden)

    H. Tjahjono

    2017-12-01

    Full Text Available Passive cooling mechanism for a nuclear reactor has been proven to be very important since the Fukushima Daiichi Reactor accident that was caused by active cooling system malfunction due to total loss of electrical power source. In the Center for Nuclear Reactor Technology and Safety of BATAN, the cooling mechanism was studied by using a natural circulation test loop named FASSIP-01 that applied thermosyphon mechanism of water inside pipes of 1” diameter. This study aimed to analytically predictthe thermal characteristics of the loop including its response time towards steady condition usingthe MATLAB calculation program. This prediction derived the influence of several parameters such as the heat transfer coefficient of the cooler side (h-cooler, the heater power, the elevation difference between the heater and cooler(DZ, and the effects of the insulation thickness of pipe (IT on the flowrate, temperature, and the heat power distribution across all components in the loop. The result showed that byavoiding boiling condition, for transferring the heater power of 1000 W and 2000 W,the needed h-cooler exceeds 200 and 400 W m-2°C-1, respectively. For a h-cooler of 200 W m-2°C-1, the circulation flow rate increased from 0.04 to 0.06 kg/s-1 for heater power increase from 1000 W to 2000 W. Those flow rates were decreased to 0.037 and 0.052 kgs-1 by increasing h-cooler to 1000 W m-2°C-1.The results were in agreement with other studies on rectangular loops in the literature.The time needed to reach 95 % towards steady state was predicted to be more than 13 hours. Reduction of this time to less than five hours was possible by reducing the heater tank volume from 100 L to 30 L or by modifying the starting heater input power.

  3. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  4. Alkali Metal Backup Cooling for Stirling Systems - Experimental Results

    Science.gov (United States)

    Schwendeman, Carl; Tarau, Calin; Anderson, William G.; Cornell, Peggy A.

    2013-01-01

    In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor. In a previous NASA SBIR Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for Stirling RPS. The operation of these VCHPs was demonstrated using Stirling heater head simulators and GPHS simulators. In the most recent effort, a sodium VCHP with a stainless steel envelope was designed, fabricated and tested at NASA Glenn Research Center (GRC) with a Stirling convertor for two concepts; one for the Advanced Stirling Radioisotope Generator (ASRG) back up cooling system and one for the Long-lived Venus Lander thermal management system. The VCHP is designed to activate and remove heat from the stopped convertor at a 19 degC temperature increase from the nominal vapor temperature. The 19 degC temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the Multi-Layer Insulation (MLI). In addition, the same backup cooling system can be applied to the Stirling convertor used for the refrigeration system of the Long-lived Venus Lander. The VCHP will allow the refrigeration system to: 1) rest during transit at a lower temperature than nominal; 2) pre-cool the modules to an even lower temperature before the entry in Venus atmosphere; 3) work at nominal temperature on Venus surface; 4) briefly stop multiple times on the Venus surface to allow scientific measurements. This paper presents the experimental

  5. Synthesis of ZnO particles in a quench-cooled flame reactor

    DEFF Research Database (Denmark)

    Hansen, Jens Peter; Jensen, Joakim Reimer; Livbjerg, Hans

    2001-01-01

    The quench cooling of a flame by injection of cold air was studied in a flame reactor for the formation of ZnO particles in a premixed flame with a precursor jet. A rapid temperature drop downstream from the temperature peak is advantageous for the attainment of a large specific surface area....... At the highest tested production rate, the specific surface area of the ZnO particles increases from 20 to 60 m(2)/g when quenching is employed. The particles are characterized by BET surface area measurements, TEM images, and the size distributions of particle aggregates are measured by a scanning mobility...

  6. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  7. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  8. Low pressure cooling seal system for a gas turbine engine

    Science.gov (United States)

    Marra, John J

    2014-04-01

    A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

  9. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  10. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  11. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR); Developpement du design d'un assemblage de controle et analyse dynamique des reacteurs a neutrons rapides de quatrieme generation refroidis au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.

    2009-07-09

    Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials. It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. During the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement - within 1{sigma} measurement uncertainty - with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal